hope creek written exam analysis draft · written exam analysis draft questions 1 through 75 are ro...

EXAM MATERIAL

Hope Creek

Written Exam Analysis DRAFT

Questions 1 through 75 are RO

Questions 76 through 100 are SRO

Question: 1 Answer: D

1 Pt(s) The Unit is operating at 100% power near the end of core life, and the OPRM system is OPERABLE. Turbine Auxiliaries Cooling System (TACS) temperatures are elevated due to high summertime ambient conditions. “A” Recirc MG Set lube oil temperature rises to 206 F, and “B’ Recirc MG Set lube oil temperature rises to 213 F.

Which ONE of the following best describes the status of the Reactor Recirc Pumps AND the appropriate operator action?

A. Both Recirc Pumps are tripped. Immediately LOCK the Mode Switch in SHUTDOWN to avoid thermal hydraulic instability.

“A” Recirc Pump is unaffected, and “B” Recirc Pump is tripped. Reduce speed on the ‘‘A” Recirc Pump to 48% in order to prevent Jet Pump vibration in the idle loop.

Both Recirc Pumps are unaffected. Adjust cooling water to restore oil temperatures within the normal band.

“A” Recirc Pump is unaffected, and “B” Recirc Pump is tripped. Reduce power by inserting control rods and monitor for power oscillations.

B.

C.

D.

Distracter Analysis:

A.

B.

Incorrect: “A” pump does not trip. Plausible: Incorrect: Pump status is correct, but reducing flow to 48% actually causes jet pump vibration in the idle loop, and reducing flow further may cause thermal hydraulic instability. Plausible: Incorrect: “B” drive motor breaker trips with lube oil temperature greater than 210 F. Plausible: Correct: A drive motor trip occurs at 2 10 F lube oil temperature. “A” is unaffected since the trip set point has not been exceeded. “B” trips since the temperature is above the set point. HC.OP-AB.RPV- 003 Condition A directs operators to insert control rods to clear the APRM Upscale alarms and monitor for power oscillations. Plausible:

C.

D.

Hope Creek 2007 ILT Exam

Level: RO Exam

Lesson Plan Objective: ????

Source: New

Level of knowledge: Analysis

Reference(s): HC.OP-AB.RPV-O03(Q), “Recirculation System 1 Power Oscillations”

KA: 295001.K1.04 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : (CFR 41.8 to 41.10) AK1.04 ?Limiting cycle oscillation: Plant-Specific 2.5 3.3


Question: 2 Answer: B

1 Pt(s) Given the following conditions:

0

0

Whch choice below describes the response of the AC Power Distribution to the conditions above?

The plant is operating at 100% power. The lOA401 bus trips due to a Bus Differential Overcurrent. All equipment operates as designed. NO operator action is taken.

A. Normal bus feeder breaker trips. Alternate bus feeder breaker closes. Motor load breakers remain closed. USS feeder breakers remain closed. EDG does NOT start

B. Normal bus feeder breaker trips. Alternate bus feeder breaker remains open Motor load breakers trip. USS feeder breakers t ip. EDG does NOT start

C. Normal bus feeder breaker trips. Alternate bus feeder breaker remains open. Motor load breakers trip. USS feeder breakers remain closed. EDG starts. EDG output breaker closes

Normal bus feeder breaker trips. Alternate bus feeder breaker trips. Motor load breakers trip. USS feeder breakers trip. EDG starts. EDG output breaker remains open

D.



A. Incorrect: Alternate feeder breaker does not close. Motor load breakers trip. USS feeder breakers trip. P 1 a u s ib 1 e :

Incorrect: USS feeder breakers trip. The EDG regular Lockout Relay prevents EDG start and prevents EDG breaker closure. Plausible: Incorrect: The EDG regular Lockout Relay prevents EDG start. Plausible:

B. Correct C.

D.

Level: RO Exam

Lesson Plan Objective: 1 EACOOE026

Source: Hope Creek Bank (Q56863)

Level of knowledge: memory

Reference(s): HC.OP-SO.PB-001 Section 3.3

KA: 295003.K2.03 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following: (CFR: 41.7 / 45.8) AK2.03 A.C. electrical distribution system 3.7 3.9



1 Pt(s) With the plant operating at rated power, the power supply fuse to a backup scram valve fails.

Which one of the following identifies the response of the associated backup scram valve and scram response due to this failure?

A.

B.

Valve repositions to trip position but NO scram occurs.

Valve CANNOT reposition but redundant valves can affect scram if an RPS hip occurs.

C. Valve CANNOT reposition and NO scram can occur even if an RPS trip occurs.

Valve repositions to trip position and a full scram occurs. D.


A.

B.

C.

Incorrect: Valve needs to energize to trip (vent). Plausible: Correct: DC solenoid valve is energize to trip (vent). There are two redundant valves, and either one can perform the trip function. Incorrect: First half is correct; however, a scram can still occur via the redundant valve. Plausible: Incorrect: Valve needs to energize to trip (vent) in ordn for a scram to occur Plausible:

D.

Level: RO Exam

Lesson Plan Objective: RPSOOOE009

Source: Hope Creek Bank (468884) Last NRC Exam 2002

Level of knowledge: Comprehension

Reference(s): HC.OP-SO.SB-0001

KA: 295003.K2.03


Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : (CFR: 41.5 / 45.6) AK3.03 Reactor SCRAM: Plant-Specific 3.1 3.5



1 Pt(s) The Unit is operating at 75% power. The “A” EHC pump is currently running with hi& vibration, and the crew is preparing to start the “B’ EHC pump to allow the “A” EHC pump to be secured for maintenance. When the “ B EHC pump is started, local operators report unusually loud noises at the EHC skid, and EHC header pressure is observed to be 1200 psig and going down. The Reactor Operator immediately places the REACTOR MODE SWITCH in SHUTDOWN, and EHC pressure stabilizes at approximately 800 psig within the next 5 seconds. Assume no other operator actions.

Which ONE of the following best describes the response of the electrical power system?

A. Generator output breakers BS2-6 AND BS6-5 automatically OPEN immediately upon placing the REACTOR MODE SWITCH in SHUTDOWN.

B. Generator output breakers BS2-6 AND BS6-5 automatically OPEN 2.5 seconds after Reverse Power Relay activation.

C. Generator output breakers BSZ-6 AND BS6-5 automatically OPEN 30 seconds after Reverse Power Relay activation.

D. Generator output breakers BS2-6 AND BS6-5 must be MANUALLY OPENED in this situation because the Reverse Power feature is bypassed with the REACTOR MODE SWITCH in SHUTDOWN.


A. Incorrect: There is no direct tie between RF’S logic and Main Turbine Generator or Output Breaker trip logic. Plausible: A reactor scram will indirectly result in a generator lockout and output breaker trip on Reverse Power as the turbine coasts down. Correct: A generator lockout will occur in 2.5 seconds after Reverse Power is sensed coincident with a turbine trip signal. In this case, a turbine trip signal came in at 1 100 psig EHC pressure. Incorrect: The 30 second time delay only comes into play if a turbine trip signal is not present. In this case, a turbine trip signal was present.

B.

C.


Plausible: A 30 second TD Reverse Power generator lockout does exist. The candidate must recognize the fact that it is only active when a turbine trip is not present. Incorrect: The breakers will open automatically without operator action due to a Reverse Power generator lockout, and there is no mode switch bypass related to the Reverse Power protective feature. Plausible: Procedures direct that the operators verify that a lockout has occurred at 0 MWe, and that is accomplished by manually opening the output breakers. In addition, the mode switch does bypass some interlocks, but reverse power is not one of them.

Level: RO Exam

D.

Lesson Plan Objective: ??

Source: New


Reference(s): NOHOlMNTURB-03, “Main Turbine Construction & Components” NOHOlMNGENOC-05, “Main Generator System” NOHOlMNPWRO-06, “Main Power” NOHOIEHCOIL-06, “EHC Control Oil”

KA: 295005.A1.07 Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP : (CFR: 41.7 / 45.6) AA1.07 A.C. electrical distribution. 3.3 3.3


Question: 5 Answer: A

1 Pt(s) Given the following:

The Unit is in OPCON 2 and a hot startup has just commenced. Reactor pressure is 250 psig. Bypass valves are controlling RPV pressure automatically. Main Turbine shell warming is in progress. The first control rod was just fully withdrawn, Turbine first stage pressure rises to 137 psig a few minutes later. The RWM Operator Display screen is on the CONFIRM SHUTDOWN screen, and is displaying the following:

ALL RODS IN: NO SHUTDOWN: NO RODS NOT FULL-IN: 00 1

Which ONE of the following best describes the current plant status?

A. The withdrawn control rod failed to scram; however, the reactor is shutdown and will remain shutdown under all conditions.

B. Plant startup is proceeding normally and the current RWM indications are also normal when the reactor mode switch is in STARTUP with all rods NOT fully inserted.

C. The withdrawn control rod failed to scram and the reactor could go critical if a cool down is initiated.

The withdrawn control rod failed to scram and the reactor is critical.

D.


A. Correct: A reactor scram signal (Main Turbine Stop Valves not full open with power greater than 30%) will be generated if Main Turbine shell warming is in service and 1" stage pressure exceeds 135.7 psig. In shell warming, the stop valves are fully closed. 30% power is sensed when 1'' stage pressure reaches 135.7 psig. Turbine procedural cautions warn operators of this fact. So, one rod failed to scram in this case. In addition, the reactor will remain shutdown under all conditions without boron with the most reactive rod fully withdrawn (shutdown margin requirement). The RWM indications


are correct for ths situation. The RWM automatically switches to the CONFIRM SHUTDOWN screen when a scram is sensed. The SHUTDOWN =NO indication is due to the way the RWM is programmed. If any one rod is withdrawn beyond 02, then the NO flag will display. The RWM lesson plan highlights this human error trap. The operator is required to understand that the reactor is and will remain shutdown with one rod fully withdrawn regardless of the RWM indication. Incorrect: The startup is NOT proceeding normally as described above, and if it was, then the RWM CONF!.RM SHUTDOWN screen should NOT be displaying. Plausible: Incorrect: The control rod did fail to scram, but the reactor should remain shutdown under all conditions without boron as discussed in the justification for answer A. Plausible: Incorrect: The control rod did fail to scram; however, the reactor should NOT be able to go critical with just one rod hlly withdrawn as discussed above. Plausible:

B.

C.

D.

Level: RO Exam

Lesson Plan Objective: ??’??

Source: New


Re ference(s): Lesson Plan NOH04RODMINC, “Rod Worth Minimizer” Lesson Plan NOHOIMNTURB, “Main Turbine Construction and Components” HC.OP-SO.AC-0001, “Main Turbine Operation”

KA: 295006.A2.05 Ability to determine andor interpret the following as they apply to SCRAM: (CFR:41.10/43.5/45.13)AA2.05Whether areactor SCRAM has occurred.4.6* 4.6*



1 Pt(s) The plant is operating at 100% power. m c k black smoke enters the Main Control Room requiring that the Control Room be abandoned. Plant equipment has not been affected by the condition. All Main Control Room actions, as specified in HC.OP-AB.HVAC.O02(Q), “Control Room Environment”, have been completed. No other actiofis have been taken.

Which ONE of the following best describes the plant parameter(s), if any, that are expected to exceed Emergency Operating Procedure entry conditions?

A. None

B. Reactor Level ONLY

C. Reactor Pressure ONLY

D. Reactor Level AND Reactor Pressure

-______-_l-__--___________________I_____----- ------_-------_--_----- Distracter Analysis:

A. Incorrect: See D below, Plausible: If the control room is abandoned before any operator action is taken, then the plant will remain in steady state operation until control is established at the RSP. In addition, an integrated level of plant knowledge is required to understand the impact that closing the MSIV’s has on feed water as well as reactor pressure control. Furthermore, knowledge of the procedure is required through training to know the control room actions taken prior to evacuation.

Plausible: Reactor level will be affected as described in D below.

Plausible: Reactor level will be affected as described in D below. Correct: The procedure directs that the MSIV’s be closed. This will result in a loss of the steam driven feed water pumps, and pressure will rise until SRV’s lift since bypass valve capability is lost when the MSIV’s are closed. Level will drop below 12.5 inches, and RPV Pressure will rise above 1037 psig, and these are both entry conditions to EOP-I01 for RPV Control.

B. Incorrect: See D below.

C. Incorrect: See D below.

D.

Level: RO Exam


Lesson Plan Objective: ABHVC2E004 Explain the reasons for how plantlsystem parameters respond when implementing Control Room Environment.

Source: New


Refmence(s): HC.OP-AB.HVAC.O02(Q), “Control Room Environment”

KA: 295016.G2.4.4 Control Room abandonment. 2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. (CFR 41.10/43.2 /45.6)



Given the following plant conditions: 1 Pt(s)

The 'A' SACS Loop is in service supplying TACS. The 'D' SACS pump is running. The 'B' SACS Pump is Cleared and Tagged. The 'A' Fuel Pool Cooling Heat Exchanger (FPCC side only) is isolated for a piping leak repair.

A lightning strike results in an 'A' channel LOCA Level 1 signal and the loss of the 10A404 4KV bus.

What is the status of the Fuel Pool Cooling System for these conditions?

A. Fuel Pool Cooling heat removal is unaffected by these conditions.

B. Fuel Pool Cooling heat removal has been lost and can be restored when the LOCA signal is reset.

Fuel Pool Cooling heat removal is being provided by the 'B' SACS Loop.

The Fuel Pool Cooling heat exchanger cross-tie valves auto open to provide Loop 'A' SACS flow to the 'B' FPCC heat exchanger.

C.

D.


A. Incorrect: The loss of the 10A404 Bus results in the loss of the ' D SACS Pump and the 'B SACS loop and the loss of cooling to the only in service FPCC HX. Plausible: Correct: The loss of the 10A404 Bus results in the loss of the 'D' SACS Pump and the ' B SACS loop. With the 'A' FPCC HX 00s on the FPCC side, all SACS cooling to FPCC is lost. The Cross-tie valve HV-2317A and HV-7922A receive close signals fiom the LOCA signal and cannot be opened without clearing the signal. The LOCA signal must be cleared to open the cross-tie valves and supply 'A' loop SACS flow to the ' B FPCC HX. Incorrect: The loss of the 1 OA404 Bus results in the loss of the 'B' SACS Loop. Plausible:

B.

C.


D. Incorrect: The valves close on a LOCA signal and have no auto open signals. Plausible:

Level: RO Exam


Source: HC Bank Q60630


Reference(s): HC.OP-SO.SM-0001

KA: 295018.K1.01 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: (CFR: 41.8 to 41.10) AKI.01 Effects on componedsystem operations 3.5 3.6


Question: 8 Answer: C

1 Pt(s) A loose fitting has resulted in a loss of instrument air to the in-service Control Rod Drive (CRD) Flow Control Valve.

Determine which of the following conditions could result kom this instrument air loss.

A.

B.

C.

D.

Control Rod Drive accumulator alarms due to low pressure.

High rod speeds during control rod withdraw.

Control Rod Drive alarms due to high temperatures.

Control Rods begin to drift due to excessive flow.

______-___________________I_____________----------------------------------------


A. Incorrect: The charging header maintains pressure on the accumulators, and the charging header taps off upstream of the FCV, so accumulator pressure will not go down. Plausible: Incorrect: The drive header is downstream of the FCV, so rod withdraw speeds will be slower rather than faster. Plausible: Correct: The FCV fails to the minimum position on a loss of air. The cooling water header is downstream of the FCV. So, the failure mode results in decreased flow to the cooling water header and temperatures will rise resulting in CRD hlgh temperature alarms. Incorrect: The failure results in low flow. The failure mode does not cause an increased DP across the dnve piston, so there is no motive force to cause the rods to drift. Plausible:

Level: RO Exam

B.

C.

D.


Source: NPO Bank (PB question)



Reference(s):

Lesson Plan NOH02000006C, “Control Rod Drive Hydraulics”

KA: 295019.K2.02 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: (CFR: 41.7 / 45.8) AK2.02 Component cooling water 2.9 3.0

P&LD M-46


Question: 9 Answer: C w 3sm- 1 Pt(s) A loss of Shutdown Cooling has occurred, and the decision is made to place

RWCU in service for alternative decay heat removal.

Which one of the following describes the appropriate action to take to maximize the heat removal rate?

A. Reduce RWCU flow through the regenerative AND non- regenerative heat exchangers by throttling total system flow.

Reduce RWCU system flow through the non-regenerative heat exchanger by bypassing the non-regenerative heat exchanger.

Reduce RWCU system return flow through the regenerative heat exchanger by bypassing the regenerative heat exchanger.

Increase RWCU system return flow through the regenerative heat exchanger.

B.

C.

D.

____-_1_______-_________I_______________-----------------------------------


A. Incorrect: Reducing total system flow will slow down the heat removal rate. Plausible: Incorrect: The non-regenerative heat exchanger is the heat sink in alternate decay heat removal mode. Reducing flow through the non- regenerative heat exchanger will slow the heat removal rate. Plausible: Correct: The regenerative heat exchanger is normally used to raise RWCU return temperature. Bypassing return flow around the regenerative heat exchanger will assist in maximizing the heat removal rate. Incorrect: Increasing return flow through the regenerative heat exchanger would slow the heat removal rate.

Level: RO Exam

B.

C.

D.


Source: New



Reference(s): HC.OP-AB.RF’V-0009, “Shutdown Cooling”

KA: 295021.K3.04 Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING: (CFR 41.5 / 45.6) AK3.04 Maximizing reactor water cleanup flow.3.3 3.4


Question: 9B Answer: C


The Unit is in OPCON 4 with RCS temperature at 175’F “B” Loop RHR is in service. “A” Loop RHR is out of service for maintenance. “A” & “ B Reactor Recirculation Pumps are out of service for maintenance. RWCU is in service with 2 pumps and 2 filters. “ B RHR pump trips due to a motor failure.

Raising RWCU total system flow will.. ...

A. increase core forced circulation.

B. prevent flashing in the RWCU suction flow venturi as the reactor heats up.

increase the reactor heat removal rate in the non-regenerative heat exchanger.

increase the reactor heat removal rate in the regenerative heat exchanger.

C.

D.

--I----- _________-__-_____--____________1_1___1_---------------~-----


A. Incorrect: Increasing RWCU flow will not increase core forced circulation inside the shroud. Plausible: Increasing RWCU flow will affect RPV flow in the annulus and bottom head area. Incorrect: Increasing flow will increase the possibility of flashing in the suction venturi. Plausible: Suction venturi flashing is a concern as RPV temperature rises and it flashing is flow dependent as well, so someone could have the misconception that increasing flow will improve the margin to flashing. Correct: Increasing RWCU system flow will increase the reactor heat removal rate in the non-regenerative heat exchanger. Incorrect: The regenerative heat exchanger is not a heat sink; it is used for reheat. Therefore, increasing system flow will not assist with reactor heat removal. The regenerative heat exchanger is fully bypassed in order to maximize the reactor heat removal rate.

B.

C.

D.


Plausible: The regenerative heat exchanger is a heat exchanger so a plausible misconception is that it will assist with reactor heat removal.

Level: RO Exam

Lesson Plan Objective: RWCUOOE003 (R) Given the necessary sheets of P&ID’s M-44 and M-45: a. Determine the normal RWCU System flowpath IAW the RWCU System Lesson Plan. b. Determine the blowdown RWCU System flowpath(s) IAW the RWCU System Lesson Plan. c. Determine the recirculation RWCU System flowpath IAW the RWCU System Lesson Plan.

Source: New


Reference(s): HC.OP-AB.RPV-0009, “Shutdown Cooling”

KA: 295021.K3.04 Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING: (CFR: 41.5 / 45.6) AK3.04 Maximizing reactor water cleanup flow.3.3 3.4




The Unit is in OPCON 5. Reactor cavity gates installed. Cavity drain down begins in preparation for reactor reassembly. Cavity level is lowered to the top of the reactor flange. 30 minutes later, operators report that fuel pool level is visibly dropping due to a significant leak through the cavity gate seals. Radiation level rise to a high of 210 m t h at the fuel storage pool area radiation monitor 95x707. Operators are able to maintain fuel pool level above the low level alarm set point.

W c h One of the following is the correct EAL declaration?

A.

B.

An EAL declaration is NOT required.

Declare an Unusual Event as a result of lowering reactor cavity level.

C.

D.

Declare an Unusual Event as a result of lowering fuel pool level.

Declare an Alert due to the potential to uncover irradiated fuel. ~ ~~


A.

B.

Correct: An EAL threshold has not been reached, so no declaration is required. Incorrect: The reactor cavity level was lowered as part of a planned and controlled evolution; therefore, the EAL is not met. Plausible: Incorrect: The fuel pool level was not intentionally lowered; however, the operators were able to control the level reduction since level is being maintained above the alarm set point. So, the EAL is not met. Plausible: Incorrect: Radiation levels are well below the EAL threshold, and the current levels in the fuel pool and reactor cavity present no significant risk of uncovering irradiated fuel. The EAL is not met. Plausible:

C.

D.


Level: RO Exam


Source: New


Reference(s): Hope Creek EALs & Bases document

KA: 295023.K2.05 Ability to operate and/or monitor the following as they apply to REFUELING ACCIDEXTS: (CFR. 41.7 / 45.6) AA2.05 ?Entry conditions of emergency plan 3.2 4.6*



A transient has occurred 10 minutes ago. The current containment conditions are as follows:

1 Pt(s)

Drywell pressure: 4.5 psig rising Drywell temperature: 150 F rising Toms pressure: 4.0 psig rising Torus water temperature: 82 F steady Torus airspace temperature: 145 F rising

Which of the following events most likely explains the current plant conditions?

A. A safety relief valve (SRV) has lifted and is discharging through a T-Quencher.

The containment is functioning normally following a feedwater line break inside containment.

B.

C. A main steam line break has occurred inside containment with a torus to drywell vacuum breaker leaking.

The containment is functioning normally following a total loss of drywell cooling.

D.


A. Incorrect: When an SRV lifts, then it normally discharges below the suppression pool water level through a T-Quencher. If this were the event in progress, then torus water temperature would be noticeably rising with a minimal increase in torus airspace temperature and pressure since the steam would be quenched by the pool. Drywell pressure and temperatures should remain normal. Plausible: Incorrect: Drywell conditions would look as indicated; however, the torus indications are wrong for this event (feedwater line break). For this event, torus water temperature should be rising as steam is forced down through the LOCA downcomers and into the pool water. Drywell pressure should be about 1.5 psig higher than torus pressure as the drywell pressure displaces the normal water column in the downcomer to push the steam into the pool. The given indications suggest that normal pressure suppression flow path has

B.


been compromised, and one of the more likely bypass paths would be through a stuck open or leaking torus to drywell vacuum breaker. Plausible: Correct: The drywell and toms airspace conditions are equalizing, but the suppression pool water temperature is normal and not rising. The condition is indicative of a high energy line break inside the drywell with a suppression pool bypass flow path, so a main steam line break inside the drywell with a leaking torus to drywell vacuum breaker could yield the dven indications. Incorrect: Drywell conditions could possibly be explained by a total loss of cooling over a long period of time; however, the given drywell pressure should not be reached within 10 minutes of a loss of cooling event. In addition, the torus conditions indicate a suppression pool bypass flow path due to the low delta P between the torus and drywell. Plausible:

C.

D.

Level: RO Exam


Source: Mod INPO Bank PB2 Question


Reference(s) : NOHOIPRICONC, “Primary Containment Structure”

KA: 295024.A2.05 Ability to determine andor interpret the following as they apply to HIGH DRYWELL PRESSURE: (CFR: 41.10 / 43.5 / 45.13) EA2.05 Suppression chamber air-space temperature: 3.6 3.7




The Unit is in OPCON 4. RPV bulk coolant temperature is 110 "F. All RPV metal temperatures are between 110 i 5 O F . The Outage Control Center has requested that RPV pressure be raised to rated pressure for hydrostatic testing while at the current RPV temperatures. The SRO responsible for the Unit denies the request.

Which one of the following is the specific reason why the SRO denied the request to raise RPV pressure under the current temperature conditions?

A.

B.

C.

The core shroud could crack due to pressurized thermal shock.

The RPV could experience brittle fracture.

SRV's could prematurely lift since the setpoints are based on hot calibration conditions.

The RPV pressure Safety Limit for OPCON 4 would be exceeded.

D.


A. Incorrect: Core shroud crachng is a BWR concern, but the failure mechanism is not related to pressurized thermal shock. Core shroud cracking is believed to be due to stress corrosion crachng. Pressurized thermal shock is more applicable to PWRs and it occurs when a component with a high delta P across it experiences a high rate temperature change. The core shroud is a vessel internal component, and the component is not subjected to any significant delta P conditions during hydrostatic testing. Furthermore, the given condition holds temperature constant. Plausible: Correct: The bases for Tech Spec 3.4.6 clearly discuss brittle fracture as the primary concern. Tech Spec 3.4.6.1 and Tech Spec Figure 3.4.6.1-1 depict the limits. Incorrect: The bases for Tech Spec 3.4.6 clearly discuss brittle fracture as the primary concern. SRV lift setpoint may vary as a function of temperature conditions, but even if it does, the brittle

B.

C.


fracture issue is the specific reason why Tech Specs forbids raising pressure to rated with the given temperature conditions, and that is specific reason why an SRO would deny such a request. Plausible: Incorrect: The RPV Pressure Safety Limit is 1325 psig and is applicable in OPCONS 1,2,3, & 4. This limit will not be approached if pressure is raised to rated conditions. Plausible:

Level: RO Exam

D.


Source: New


Reference( s): Tech Spec 3.4.6 &Bases NOHOIRXVESSC. “Reactor Vessel and Intemals”

KA: 295025.G2.1.32 2.1.32 Ability to explain and apply all system limits and precautions. (CFR: 41.10 / 43.2 / 45.12)



1 Pt(s) The reactor has scrammed on high drywell pressure. Plant conditions are as follows:

Reactor pressure: RF'V water level:

SO psig, stable -50 inches, rising slowly

Suppression pool temperature: 200 "F, rising slowly Suppression chamber pressure: 10 psig, stable Suppression pool water level: 0 inches, stable ''E' RHR is in Toms Cooling and Spray: 10,000 gpm

For the conditions stated, which of the following is the highest suppression pool temperature that ensures NPSH is maintained for the "B" RHR pump?

A. 206 "F (* 1 "F)

B.

C.

D.

223 "F (=k 1 "F)

233 "F (* 1 "F)

246 "F (5 1 "F)


A.

B.

C.

D.

Incorrect: Based on overpressures of 0 psig. Plausible: Incorrect: Based on overpressures of S psig. Plausible: Correct: EOP Caution 2 based on 10 psig suppression chamber overpressure curve. Incorrect: Based on overpressures of 15 psig. Plausible:

Level: RO Exam





Reference( s): EOP Caution 2

KA: 295026.K1.01 Knowledge of the operational implications of rhe following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: (CFR: 41.8 to 41.10) EKI.01 Pump NPSH 3.0 3.4



Given the following conditions: 1 Pt(s)

The plant is at 100% power. A loss of drywell cooling causes elevated drywell temperature and pressure.

Which statement below describes Narrow Range Level Instrument response to the conditions above?

A.

B.

C.

D.

Because both reference and variable leg densities increase, there will be NO ehange in indicated level.

Because narrow range instruments are density compensated, there is NO change in indicated level.

Inereased reference leg density will cause a decrease in indicated level.

Decreased reference leg density will cause an increase in indicated level.


A. Incorrect: Both densities will decrease, but the reference leg has more vertical piping that is subject to the high temperature conditions. Plausible: Incorrect: The reference leg has more vertical piping that is subject to the high temperature condition thereby causing indicated level to rise. Plausible: Incorrect: %sing drywell temperature will lower reference leg density causing indicated level to rise. Plausible: Correct: Level detectors measure the DP between the reference leg (Hi press) and the variable leg (Low press). With actual RPV level high, the Dp is at its smallest value. Any condition causing the pressure exerted by the reference leg to drop or cause the pressure exerted by the variable leg to rise will cause indicated level to rise or be higher than actual. The converse also applies.

B.

C .

D.


Level: RO Exam

Lesson Plan Objective: RXINSTEO 1 8 INSTREEOOI

Source: HC Bank Q54798 Last used on 9/13/99


Reference(s) :

KA: 295028.K2.02 Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: (CFR: 41.7 145.8) EK2.02 Components internal to the drywell 3.2 3.3



A suppression pool level transient is in progress and suppression pool level reaches 54 inches and continues to lower. A manual reactor scram is inserted in accordance with the EOP’s.

1 Pt(s)

What is the basis for this action?

A. The SRV tailpipe vacuum breakers are uncovered, so the containment may be directly pressurized if the SRV’s lift.

The torus vent header drain lines are almost uncovered. B.

C.

D.

Tech Specs require an immediate reactor scram.

Low Pressure ECCS is inoperable and unavailable due to NPSH and vortex concerns.


A. Incorrect: The SRV tailpipe vacuum breakers are located in the drywell and are supposed to be uncovered to allow proper operation once an SRV closes and steam in the tailpipe condenses. Also, SRV lift should not cause the vacuum breaker to open while the SRV is actively discharging. Plausible: SRV tailpipe vacuum breakers are normally uncovered. A failed open or leaking vacuum breaker could result in direct containment pressurization upon SRV lift, but the stem of the question does not suggest that there are any problems with the vacuum breaker. The distracter tests the candidates’ knowledge of the location and operation of the vacuum breaker. Correct: The EOP Basis document provides several reasons for directing a manual scram as a prudent anticipatory action. The most immediate reason is that if suppression pool level cannot be maintained above 55 inches, then the torus vent header 1 56 inch drain lines are close to becoming uncovered (uncovered at 50 inches). These drain lines are not capped or valved. If these lines are uncovered, then the SP airspace could be directly pressurized in the event of a LOCA since some steam would not be quenched by the

Incorrect: LCO 3.5.3.a Action A requires that level be restored to at least 74.5 inches within 1 hour or be in hot shutdown within the next 12 hours. An immediate scram is not required for this condition.

B.

pool. C.


Plausible: The EOP bases state that that a manual scram is a prudent action to take since efforts to restore level above the LCO requirement have been unsuccessful and that further level degradation is likely to cause the torus vent drain lines to become uncovered. Incorrect: Low pressure ECCS remains available despite potential operability concerns related to NPSH & vortexing. In addition, the EOP bases do NOT list this as a basis to initiate a manual scram. Plausible: The Tech Spec LCO 3.5.3 bases mention that NPSH and vortex issues are a concern as level drops below 74.5 inches, but that is not the reason that the EOP’s direct a manual scram.

D.

Level: RO Exam


Source: New

Level of knowledge: Memory

Reference(s): Tech Spec 5.5.3 &Bases EOP 102 & Bases

KA: 295030.K3.06 Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL. (CFR. 41.5 i 45.6) EK3.06 Reactor SCRAM. 3.6 3.8




A small break LOCA is in progress. All control rods are hlly inserted. All injection sources have failed. Reactor level is - 170 inches and going down slowly. Reactor pressure is 900 psig and trending down very slowly. Operators just began to perform OP-EO.ZZ-3 10, “Alternate Injection Using Fire Water”, but those actions are not yet complete.

When must the operators open 5 ADS valves?

A.

B.

BEFORE RF’V level drops to - 185 inches.

WHEN RPV level CANNOT be restored AND maintained above - 185 inches.

C. BEFORE RF’V level drops to - 200 inches.

D. WHEN RPV level drops to - 200 inches.


A. Incorrect: An emergency blowdown is allowed by EOP-101 before - 185 is reached if sufficient injection sources are available to restore level above - 129 inches or restore and maintain it above - 185 in following the blowdown. Ln the scenario given, there is zero injection available. Plausible: Incorrect: If level cannot be restored and maintained above - 185 inches with an injection source available, then SAG entry is required for containment flooding. With no injection lined up and ready to inject, EOP-101 directs the use of steam cooling down to -200 inches before an emergency blowdown is directed. Plausible: Incorrect: See explanation for D. With zero injection capability, it is not appropriate to blowdown until - 200 inches is reached since steam cooling should be used as long as possible until it is no longer effective to maintain adequate core cooling. Plausible:

B.

C.


D. Correct: With zero injection, EOP-101 directs that the operators wait until the Minimum Zero Injection Water Level is reached (MZIRWL, - 200 inches). This allows steam cooling to be used for as long as possible before an emergency depressurization is performed. Once MZIRWL is reached, then adequate core cooling is not assured and an emergency depressurization is required even without an injection source available.

Level: RO Exam


Source: New


Reference(s) : EOP-IO1 and EOP-202

KA: 295031.A1.07 Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.7 145.6) EAI.07 Safety/relief valves 3.7* 3.7*



1 Pt(s) The Unit was operating at 100% power when the following event occurred:

0

0

0

The reactor scrammed due to high reactor pressure. Reactor pressure peaked at 1 1 15 psig during the transient, and pressure is now at 900 psig and lowering slowly. 5 control rods still indicate position 48. All APRM downscale lights are extinguished. One SRV is stuck open. Suppression pool temperature is 112 "F and rising. Reactor level is at - 25 inches and lowering slowly. Drywell pressure is 0.5 psig and stable.

Which ONE of the following is the correct operator action to take based on these conditions?

A. Maintain RPV level between - 25 inches and + 54 inches.

B.

C.

Lower RPV level until UNTIL level drops below - 50 inches.

Lower RF'V level UNTIL reactor power drops below 4% RPV level reaches - 129 inches OR all SRV's are closed.

D. Maintain RPV level between -185 inches and - 25 inches.


A. Incorrect: With an ATWS in progress with power above 4%, maintaining the current reactor level or allowing it to rise will allow energy addition to the containment to continue. In fact, raising level will tend to raise power making the situation worse. Plausible: Incorrect: This answer would he correct if suppression pool temperature was less than 100 "F, but it is not, so answer Cis the correct action to take. Plausible: Correct: Answer C is correct since reactor power is above 4% (APRM downscale lights extinguished) and suppression pool temperature is above 1 10 "F with an SRV open. Incorrect: Additional level reduction is required to either get power less than 4% or the SRV closed to help minimize containment energy addition. Once that is accomplished, then a maintenance

B.

C.

D.


band is appropriate between - 185 inches and the level needed to keep power less than 4 % or to get the SRV to close or - 129 inches (if power does not drop to less than 4% or the SRV does not close). Plausible:

Level: RO Exam


Source: New


Reference( s) : EOP-lO1A & bases

KA: 295037.A2.04 Ability to determine andor interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : (CFR: 41.10 / 43.5 / 45.13) EA2.04 Suppression pool temperature 4.0* 4.1*




A main steam line break has occurred in the turbine building immediately upstream of the main turbine stop valves. The break CANNOT be isolated. Steam is discharging into the turbine building. Health Physics reports that radiation levels are rising in the turbine building as well as outside the turbine building. The Emergency Director just declared a Site Area Emergency.

0

Who should be notified as the first priority?

A, The states and counties.

B. The NRC Operations Center.

C.

D. On site plant personnel.

The Site Vice President or designee.


A. Incorrect: The states & counties must be notified within the next 15 minutes to comply with regulations and ECG Attachment 6. This would be the second highest priority notification to make. The most immediate priority is to protect the health & safety of people that are on site, so that notification should be made first. Plausible: Incorrect: The NRC Ops Center must be notified withm 60 minutes, so this is not the first notification that needs to he made. Plausible: Incorrect: There is no time requirement to make this notification even though it may be an important notification to make. It may be made as soon as practical and it is not the first priority. Plausible: Correct: A plant page notification should be made immediately to warn plant personnel of the ongoing hazard. This action is required to protect the health and safety of people, and it is the number 1 priority. Other notifications still need to be made within specified time requirements, but this is the most immediate concern and notification should not be delayed.

B.

C .

D.


Level: RO Exam


Source: New


Reference(s): ECG Attachment 6 OP-AA-104-101, “Communications”

KA: 295038.G2.1.14 2.1.14 Knowledge of system status criteria which require the notificationofplant personnel. (CFR: 43.5 / 45.12)



1 Pt(s) An electrical fire has been reported in the “A” Emergency Diesel Generator (EDG) room. The automatic fire suppression system actuated as designed. Fire response personnel are planning to enter the room to confirm that the fire is extinguished.

Which one of the following describes the fire class and the potential hazards associated with entering the room?

A.

B.

C.

D.

Class C; suffocation hazard due to C 0 2 discharge.

Class C; electrocution hazard due to water deluge.

Class B; suffocation hazard due to C02 discharge.

Class B; electrocution hazard due to water deluge.


A. Correct: The fire is electrical in nature, so it is Class C. The EDG room is protected by automatic C02 suppression, so suffocation is a major hazard to consider prior to entry. Incorrect: Class C is correct; however, the room is NOT protected by water deluge. Plausible: Incorrect: Class B pertains to combustible liquids and this fire was electrical. The second part of the answer is correct. Plausible: Incorrect: Class B pertains to combustible liquids and this fire was electrical. In addition, the room is NOT protected by water deluge. Plausible:

B.

C.

D.

Level: RO Exam


Source: INPO Bank (Dresden 2)


Reference( s): NOHO4EDGOOOC, “Emergency Diesel Generator”


NOHOIFIREPRO, “Fire Protection System”

KA: 600000.K1.01 Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: AKI.01 Fire Classifications by type 2.5 2.8



1 Pt(s) The Unit is in OPCON 1 with irradiated fuel moves in progress on the refueling floor. Ventilation systems are lined up in the normal lineup to support the plant conditions. The following event occurs:

An irradiated Fuel bundle is dropped from the full up position. The bundle impacts other irradiated bundles in the fuel pool. Bubbles are observed rising from the dropped bundle. Refuel Floor Exhaust radiation rises to 1.8 X 10E-3 pCi/cc. Reactor Building Exhaust radiation is normal. 60 seconds later the Reactor Operator checks ventilation status.

Which of the following describes secondary containment ventilation status?

A. Reactor Building ventilation supply and exhaust fans are running, FRVS vent fans are NOT running, and FRVS recirculation fans are NOT running.

Reactor Building ventilation supply and exhaust fans are NOT running, ONE FRVS vent fan is running, and SIX FRVS recirculation fans are running.

B.

C. Reactor Building ventilation supply and exhaust fans are NOT running, TWO FRVS vent fan are running, and SIX FRVS recirculation fans are running.

Reactor Building ventilation supply and exhaust fans are running, FRVS vent fans are NOT running, and FOUR FRVS recirculation fans are running.

D.


A. Correct: The Refuel Floor Exhaust high radiation isolation signal occurs at 2.0 X 10E-3 pCi/cc. In the given scenario, Refuel Floor Exhaust radiation levels only reached 1.8 X 10E-3 pCi/cc; therefore, plant ventilation system status remains unaffected. RBVS supply & exhaust fans are still running, and FRVS fans are not running. Incorrect: This answer would be correct if the radiation levels exceeded the isolation setpoint, but the radiation levels are below the setpoint in this scenario, so this answer is not correct. Plausible:

B.


C. Incorrect: Answer C is partially correct when the radiation level exceeds the setpoint; however, only the lead FRVS vent fan would auto start on the initiation signal. The second FRVS vent fan will only automatically start if there is an initiation signal and the lead fan has low flow for 45 seconds. Plausible: Incorrect: Answer D is partially correct for a normal ventilation lineup; however, the FRVS recirculation fans are not pat of the normal ventilation lineup. Plausible:

D.

Level: RO Exam


Source: New


ReferenMs): HC.OP-AB.CONT-0005, “Irradiated Fuel Damage” Lesson Plan NOHOlRBVENTC, “Reactor Building Ventilation”

KA: 295023.K2.05 Knowledge of the interrelations between REFUELING ACCIDENTS and the following: (CFR: 41.7 / 45.8) AK2.05 Secondary containment ventilation 3.5 3.7



1 Pt(s) The Unit was operating at 100% power when a total loss of feed water occurred. HPCI and RCIC automatically started and RPV level rose to +56 inches.

which of the following describes the correct valve status and operation?

A. HV-4282, “RCIC Turbine Trip and Throttle Valve” is closed. The valve can ONLY be opened after the high level condition clears AND the operator manually resets the valve trip latch by using the control room close pushbutton to reengage the valve operator to the valve stem. Once this is done, then the operator can open the valve using the open pushbutton.

HV-4282, “RClC Turbine Trip and Throttle Valve” is closed. If RPV level drops to - 38 inches, then the valve will open automatically without operator action.

HV-F045, “RCIC Turbine Steam Supply Isolation Valve” is closed. If RPV level drops to - 38 inches, then the valve will open automatically without operator action.

HV-F045, “RCIC Turbine Steam Supply lsolation Valve” is closed. The valve can ONLY he opened after the high level condition clears AND the operator manually resets the valve trip latch by using the control room close pushhutton to reengage the valve operator to the valve stem. Once this is done, then the operator can open the valve using the open pushbutton.

8.

C.

D.

________________________l__________l____----------------------------


A. Incorrect: HV-4282 does not trip on a hgh level shutdown signal. The operator action described is correct to reset the valve if it tripped; however, the valve should not have tripped in the given situation. Plausible: Incorrect: HV-4282 does not trip on ahigh level shutdown signal. Plausible: Correct: The valve closes on a +54 inch siaal . The valve will automatically reopen on -38 inches. The system is designed this way to allow the system to cycle between -38 inches and + 54 inches without operator intervention.

B.

C.


D. Incorrect: The valve status is correct, but the operation describes the reset procedure for the HV-4282 valve. Plausible:

Level: RO Exam


Source: New


Reference(s): Lesson Plan HOHORCICOO, “RCIC System”

KA: 295008.K3.08 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR WATER LEVEL: (CFR: 41.5 145.6) AK3.08 RCIC steam supply valve closure: Plant-Specific.3.4 3.5




The Unit is operating at 100Y0 power. A feed water malfunction causes 2 Reactor Feed Pumps to trip and the remaining feed pump to run down to minimum speed. RPV level briefly dipped below - 38 inches for about 5 seconds before operators were able to reestablish feed flow. RPV is now - 25 inches and rising slowly. All control rods fully inserted as rcquired.

What is the status of the Reactor Recirculation Pumps 30 seconds later?

A.

B.

C.

D. Both pumps are tripped.

Both pumps are running at the pre-transient speed.

Both pumps are running at 45% speed.

Both pumps are running at 30% speed.


A.

B.

C.

Incorrect: Both pumps tripped 9 seconds after - 38 inches was reached. Incorrect: The 45% speed limiter was activated; however, both pumps tripped 9 seconds after - 38 inches was reached. Incorrect: The 30% speed limiter was activated as level dropped below + 12.5 inches during the transient; however, both pumps tripped 9 seconds after - 38 inches was reached. Correct: Both pumps trip 9 seconds after - 38 inches is reached. The signal seals in. so the trip still occurs even if level recovers within the 9 second time delay.

D.

Level: RO Exam


Source: New



Reference( s): Lesson Plan NOHOlRECCON, “Reactor Recirc Flow Control” Lesson Plan NOHOIRECIRC, “Reactor Recirculation System”

KA: 295009.A1.03 Ability to operate and/or monitor the following as they apply to LOW REACTOR WATER LEVEL: (CFR: 41.7 145.6) AA1.03 Recirculation system: Plant-Specific. 3.0 3.1




The plant is operating at 100% power. The DRYWELL PRESSURE HVLO alarm is received. Drywell pressure is 1 .I 1 psig. HC.OP-AB.CONT-001, D w e l l Pressure abnormal is entered.

Which of the following events, by itself, could be the cause of the pressure rise?

A.

B.

C.

D.

Failure of the "A" Reactor Recirculation Pump #2 Seal

FV-4971 Nitrogen Flow Control Valve fails open

Torus Vent Valve Isolation Valve HV-11541 fails open

Loss of power to multiple Drywell Fans


A.

B.

C.

D.

Incorrect: Assuming the #I seal is intact, no change in DW conditions will occur. Plausible: Incorrect: During normal operation, the nitrogen FCV is isolated from the DW. Plausible: Incorrect: A vent valve opening would result in a reduction in pressure. However the rupture disk downstream of HV-11541 should be intact resulting in no effect on DW pressure Plausible: Correct: A reduction in cooling will raise temperature and therefore pressure in the Drywell.

Level: RO Exam

Lesson Plan Objective: ABCNTlE004

Source: HC Bank 461761



Reference( s) : HC.OP-AB.CONT-0001

KA: 295012.A2.02 Ability to determine and/or interpret the following as they apply to HIGH DRYWELLTEMPERATURE: (CFR: 41.10 / 43.5 / 45.13) AA2.02 Drywell pressure. 3.9 4.1



1 Pt(s) Consider each of the following scenarios independently:

1. A Site Area Emergency was declared based on high FRVS release rate indications exceeding the EAL threshold.

2. An Alert was declared based on high SPV release rate indications exceeding the EAL threshold.

3. An Alert was declared based on a large radioactive liquid release to the river.

Select the answer below that provides a complete list of the scenarios which satisfy the entry conditions for HC.OP-EO.ZZ-0104, “Radioactive Release Control”.

A. 1and2and3

B. 2and3

C. 1 and2

D. 1 only


A. Incorrect: EOP-104 is entered based on any gaseous radioactive release above the Alert EAL threshold. Scenario 3 deals with a liquid effluent release. Plausible: Scenarios 1 & 2 do meet the entry conditions for EOP- 104. Scenario 3 partially meets the entry conditions since it is a radioactive effluent release, there is an Alert EAL for this type of release, and finally, the EOP is entered at the Alert EAL level. However, Scenario 3 is NOT for a gaseous effluent release. Incorrect: EOP-104 is entered based on any gaseous radioactive release above the Alert EAL threshold. Scenario 3 deals with a liquid effluent release. Plausible: Scenario 2 does meet the entry conditions. Scenario 3 partially meets the entry conditions since it is a radioactive effluent release, there is an Alert EAL for this type of release, and finally, the EOP is entered at the Alert EAL level. However, Scenario 3 is NOT for a gaseous effluent release.

B.


C. Correct: EOP-104 is entered based on any gaseous radioactive release above the Alert EAL threshold. Scenarios 1 AND 2 satisfies the entry requirements. Incorrect: Scenario 2 is also a valid entry condition for EOP-104, so Scenario 1 alone is not correct. Plausible: The Site Area Emergency EAL threshold could be mistaken as the EOP-104 entry threshold, and if it was, then the answer would be correct since Scenario 2 would be eliminated.

D.

Level: RO Exam

Lesson Plan Objective: EOP103E002 Given a set of plant conditions, analyze and determine if entry conditions into HC.OP-EO.ZZ-O103/4 exists.

Source: New


Reference(s) : HC.OP-EO.ZZ-0104 &Bases

KA: 295017.G2.4.4 2.4.4 Ability to recognize abnormal indications for system operating parameters whch are entry-level conditions for emergency and abnormal operating procedures. (CFR 41 .IO / 43.2 / 45.6)




The Unit is critical in OPCON 2. Reactor pressure is 550 psig. “A” CRD pump is out of service for maintenance. “B” CRD pump tripped at 1O:OO EDT due to 3 motor fault. At 10:05 EDT, at least 5 different CRD accumulator trouble alarms were received due to low accumulator pressure. The associated control rods are fully withdrawn.

Which one of the following describes the current control rod insertion capability?

A. Manual insertion capability is NOT available using RMCS. Scram function is fully assured.

Manual insertion capability is available using RMCS. Scram function may be degraded.

Manual insertion capability is NOT available using RMCS. Scram function may be degraded.

Manual insertion capability is available using RMCS. Scram function is fully assured.

B.

C.

D.


A. Incorrect: The first half is correct since RMCS is not available, however scram function is not fully assured. See explanation for answer C. Incorrect: The first half is incorrect since RMCS is not available without drive pressure. The second half is correct since scram function may be degraded. See explanation for answer C. Correct: With both CRD pumps off, there is no drive pressure, so manual insertion using FUvlCS is not possible. Furthermore, CRD charging water pressure is also lost. At least 5 withdrawn control rods have low accumulator pressures rendering them inoperable. In addition, reactor pressure is less than 900 psig, so reactor pressure alone may not complete the scram. Rods may or may not fully insert or they may not insert fast enough under the given conditions, so the scram function may be degraded.

B.

C.


D. Incorrect: Both parts of this answer are wrong. See the explanation for answer C.

Level: RO Exam


Source: Modified NPO Bank (Cooper Question)


Reference(s): HC.OP--.IC-0001, “Control Rod” Tech Spec LCO 3.1.3.5 and Bases

KA: 295022.K1.01 Knowledge of the operational implications of the following concepts as they applyto LOSS OF CRDPUMPS: (CFR: 41.8 to 41.10) AKI .01 Reactor pressure vs. rod insertion capability. 3.3 3.4

HOE Creek 2007 ILT Exam


1 Pt(s) A Design Basis LOCA has occurred. ALL Reactor Building Area Radiation Monitors are upscale.

%ch of the following describes the response of the AMs '?

A. The response is NOT expected. Tbe detectors are saturated.

B.

C.

The response is expected. The detectors are saturated.

The response is NOT expected. The ARMs failed safe due to power interruption.

The response is expected. The ARMs failed safe due to power interruption.


D.

______-___-_-_______l_l_________________------------------------------------

A. Incorrect: The response is EXPECTED for a DBA LOCA. During post accident scenarios, all Reactor Building A M s will be upscale and will remain so for about 30 days after the event. Plausible: The second half of the answer is correct. The answer could be selected if a candidate is not familiar with post accident secondary containment rad conditions. They may think that rad levels do not rise to the level to drive the ARMs upscale. Correct: During post accident scenarios, all Reactor Building ARMs will be upscale and will remain so for about 30 days after the event. Incorrect: The response is EXPECTED for a DBA LOCA. During post accident scenarios, all Reactor Building ARMS will be upscale and will remain so for about 30 days after the event. In addition, the ARMs do not fail upscale on power loss. Plausible: The answer could be selected if a candidate is not familiar with post accident secondary containment rad conditions. They may think that rad levels do not rise to the level to drive the ARMs upscale. In addition, the answer might be selected by someone believing that power loss could lead to an upscale failure (fail safe). Incorrect: The first half is correct in that the response is expected, but the response is not due to power loss. Plausible: The answer could be selected if a candidate is not familiar with post accident secondary containment rad conditions,

B.

C .

D.


especially if they believed that power loss could lead to an upscale failure (fail safe).

Level: RO Exam

Lesson Plan Objective: RADMONE002 From memory, briefly explain how plant radiation levels can be used to determine the extent of core damage.

Source: HC Bank (Audit Exam 1999) Note: Slightly modified answer sentence structure and wording to improve plausibility. This is still considered a bank question.


Reference(s): Lesson Plan NOHOIRADMON, “Process and Area Radiation Monitoring Response”

KA: 295033.K2.02 Knowledge of the interrelations between HIGH SECONDARY CONTAINMENT AREA RADLATION LEVELS and the following: (CFR: 41.7 / 45.8) EK2.02 Process radiation monitoring system. 3.8 4.1



1 Pt(s) The Unit is operating at 100% power with RCIC out of service for maintenance. A large seismic event occurs resulting in the following:

A loss of offsite power has occurred and all EDG’s are powering their respective buses. All control rods are fully inserted. HPCI is running and controlling RF’V level at + 25 inches. Suppression pool level is 69 inches and lowering due to a leak on the HPCI pump suction line. RWCU Pipe Chase temperature is above Max Safe Operating Temperature and rising due to a leak whch CANNOT be isolated. The HPCI room water level is above the Max Safe Operating Floor Level and rising. The RCIC room water level is above the Max Safe Operating Floor Level and rising. RPV pressure is 900 psig and stable. An Emergency Depressurization is ordered.

Which of the following describes the reason for the Emergency Depressurization?

A. HPCI must be secured with room water level above the Max Safe Operating water level; therefore, low pressure ECCS systems will be required for level control.

A primary reactor coolant system is discharging into the reactor building with the HPCI & RCIC rooms above the Max Safe Operating water level.

The primary containment system is discharging into the reactor building with the HPCI & RCIC rooms above the Max Safe Operating water level.

The primary containment & a primary reactor coolant system are both discharging into the reactor building with the RWCU Pipe Chase above the Max Safe Operating temperature.

B.

C.

D.



A. Incorrect: Even if HPCI did need to be secured, then an emergency depressurization would not be required at this point since RPV level is at + 25 inches, and SLC is still available for high pressure injection. Even if there were a total loss of all hgh pressure injection sources, then an emergency depressurization would not be required until level reached - 185 inches. Correct: EOP-103 requires an emergency depressurization when a primary coolant system is discharging into the reactor building and more than one area exceeds the Max Safe Operating level for the same parameter. In this case RWCU (part of the RCS) is actively discharging into the reactor building. In addition, The Max Safe Operating level has been exceeded in two different areas (HPCI room & RCIC room). Incorrect: The leak in primary containment system leak is the cause of the high water level in the HPCI & RCIC rooms; however, the potential primary containment problem is not a direct reason for the emergency depressurization in EOP-103. The combination of high water level in the HPCI & RCIC rooms combined with the RWCU leak (primruy system discharging into the reactor building) is the reason for the emergency depressurization. Incorrect: One area above Max Safe Operating temperature is not a sufficient reason to perform an emergency depressurization. See justification for answer B for the requirements for an emergency depressurization.

B.

C.

D.

Level: RO Exam


Source: New


Reference(s): EOP-103 & Bases

KA: 295036.K3.01 Knowledge of the reasons for the following responses as they apply to SECONDARY CONTANMENT HIGH SUMP/AREA WATER LEVEL : (CFR 41.5 / 45.6) EK3.01 Emergency depressurization. 2.6 2.8




All ECCS is injecting.

A large break LOCA occurred 2 minutes ago. W V pressure is 125 psig and dropping.

RPV level is - 145 inches and rising slowly. The “B” RHR pump is exhlbiting symptoms of suction strainer clogging and may have inadequate NPSH.

Which of the following describes the correct action to take at this time?

A. Throttle HV-F003B, “B RHR HX Shell Side Outlet MOV” to reduce “B” LPCI injection flow.

B. Throttle HV-F048B, “RHR HX B Shell Side Byp MOV” to reduce “B” LPCI injection flow.

C.

D.

Shift “B” RHR suction from the suppression pool to the CST.

Maximize all ECCS injection flow until RPV level is restored above + 12.5 inches.


A. Correct: HC.OP-AB-ZZ-0155 provides an option to throttle flow using HV-F003B & HV-FO48 as needed to maintain NPSH. This action is permitted as long as RPV level can be maintained above TAF. In this case, level is above TAF and trending up indicating that there is margin to reduce injection flow. The procedure alerts the operator that the HV-F048 is interlocked such that it cannot be closed within the first 3 minutes of a LOCA. So, throttling HV- F003B is the correct answer. Incorrect: HV-F048 is not available to be throttled at this time since it has only been 2 minutes since the LOCA signal came in. Valve closure is locked out for 3 minutes. Incorrect: HC.OP-AB-ZZ-0155 provides this as an option for “B’ Core Spray, but there is no CST suction path for RHR. Incorrect: If RPV level cannot be maintained above TAF, then ECCS injection flow must be maximized regardless of NPDH; however, in the given scenario, RPV level is well above TAF and is

B.

C.

D.


rising indicating that there is margin to reduce flow on “A” LPCI without jeopardizing adequate core cooling.

Level: RO Exam


Source: New


Reference(s) : HC.OP-AB-ZZ-0155, “Degraded ECCS Performance / Loss of NPSH’

KA: 203000.A2.01 Ability to (a) predict the impacts of the following on the RHFULPCI: INJECTION MODE (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) A2.01 Inadequate net positive suction head 3.2 3.4



RHR Loop B is operating in the Shutdown Cooling (SDC) Mode, when RPV pressure rises to 83 psig.

1 Pt(s)

Which one of the following describes the expected system response? (Assume all systems being operated fiom the control room.)

A. RHR pump suction valve (F006B) auto closes, RHR Pump ‘B’ trips, and shutdown cooling return isolation valve (F015B) auto closes

B. Shutdown cooling suction valves (FOO8 and F009) auto close, RHR Pump ‘B’ trips, and shutdown cooling return isolation valve (F015B) auto closes

C. Shutdown cooling suction valves (F008 and F009) auto close, FLHR pump suction valve (F006B) auto closes, and RHR Pump ‘B’ trips

RHR Pump ‘B’ trips, min flow control valve (F007B) auto opens, and shutdown cooling return isolation valve (F015B) auto closes.


D.

A.

B. C.

D.

Incorrect: There is no auto closure interlock for F006B, at 82# (rising): F008F009 close; Pump trips; F015 closes. Correct: A SDC isolation has occurred at X32 psig. Incorrect: There is no auto closure interlock for F006B, at 82# (rising): F008iF009 close; Pump trips; F015 closes. Incorrect: F007 enabled only if pump breaker is closed, at 82# (rising): F008iF009 close; Pump trips; F015 closes

Level: RO Exam

Lesson Plan Objective: RHRSYSEOl 1

Source: HC Bank (Q56406)


Reference(s) : HC.OP-SO.BC-0002


KA: 205000.A3.01 Ability to monitor automatic operations of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) including: (CFR: 41.7/45.7)A3.01 Valveoperation 3.2 3.1



1 Pt(s) The Unit was operating at 100% power when a Group I isolation occurred. HPCI status is as follows:

0

HPCI is operating in the pressure control mode. The HPCI flow controller FIC-R600 is in AUTOMATIC and controlling flow at 3000 GPM. The HPCI TURBINE TROUBLE alarm was just received due to turbine oil cooler high temperature. HPCI turbine bearing temperatures are reading 155°F and rising slowly on Control Room recorder TR-R605, The Reactor Operator identifies that HPCI speed is 2000 RPM.

0

Which one of the following actions should be taken to attempt to correct the hgh temperature conditions so that HPCI can remain in service?

A. Throttle HV-F008, “TEST BYP TO CST MOV” in the OPEN direction to RAISE turbine speed.

Throttle HV-F008 “TEST BYP TO CST MOV” in the OPEN direction to LOWER turbine speed.

Throttle HV-F008 “TEST BYP TO CST MOV” in the CLOSED direction to RAISE turbine speed.

B.

C.

D. Throttle HV-FOOS “TEST BYP TO CST MOV” in the CLOSED direction to LOWER turbine speed.

1___-_-_---__-_1____-------------------------------------------------------------


A. Incorrect: Throttling the valve open will lower flow resistance; therefore, turbine speed will lower as the FIC attempts to maintain the 3000 gpm set point. Lowering speed is a non-conservative action which will make the temperature problem worse and could lead to turbine bearing failure. See explanation for answer C. Incorrect: Opening the HV-F008 valve will lower speed, but the correct action is to raise turbine speed above 2150 RPM. See explanation for answer C. Correct: HC.OP-SO.BJ-0001, “High Pressure Coolant Injection System Operation” precaution 3.1.1 warns the operator to minimize HPCI turbine operations less when turbine speed is less to 2150 RPM in order to prevent turbine bearing or exhaust check valve

B.

C.

How Creek 2007 ILT Exam

chatter, so turbine speed needs to be raised in this scenario. Throttling the CST return valve in the closed direction increases the resistance to flow. The FIC is set to automatically control flow at 3000 gpm; therefore, the controller will call for an increase in turbine speed to maintain 3000 gpm. The increase in turbine speed will also increase the speed of the turbine driven oil pump which should resolve the bearing high temperature and oil cooler temperature. So, answer C is correct. Incorrect: Closing HV-F008 will cause turbine speed to rise not lower. See explanation for answer C.

D.

Level: RO Exam

Lesson Plan Objective: HPCIOOE007 From memory, state the bases for HPCI turbine speed limitations during operation IAW HC.OP-SO.BJ-0001

Source: New


Reference(s) :

0 HC.OP-ABZZ-0001, “Transient Plant Conditions”

HC.OP-SO.BJ-0001, “High Pressure Coolant Injection System Operation”

NOHOIHPCIOO, “High Pressure Coolant Injection System”

KA: 206000.A4.03 Ability to manually operate andor monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) A4.03 Turbine temperatures: BWR-2,3,4 3.1 3.0




The Unit is in OPCON I . ECCS Jockey Pump 1DP228 tripped. The CORE SPRAY LOOP B TROUBLE alarm was received due to low injection line pressure.

Which system should be lined up to clear the alarm condition, and where does this system inject into the Core Spray loop?

A. DEMIN WATER, which injects DOWNSTREAM of HV-FOOSB, “Core Spray Loop Inboard Isolation Valve”.

DEMIN WATER, which injects UPSTREAM of HV-FOOSB, “Core Spray Loop Inboard Isolation Valve”.

CONDENSATE TRANSFER, which injects DOWNSTREAM of HV-FOOSB, “Core Spray Loop Inboard Isolation Valve”.

CONDENSATE TRANSFER, which injects UPSTREAM of HV-FOOSB, ‘‘Core Spray Loop Inboard Isolation Valve”.

B.

C.

D.

_____________--_--_____1_1_______1_1____---------------------------------


A. Incorrect: Demin water is not correct. Piping is installed from condensate transfer to each Core Spray loop, and a normally closed manual valve can be opened to provide keep fill to Core Spray from condensate transfer. Incorrect: Demin water is not correct. Piping is installed from condensate transfer to each Core Spray loop, and a normally closed manual valve can be opened to provide keep fill to Core Spray from condensate transfer. Incorrect: This is the correct system; however the injection point is upstream of HV-FOO5B.

B.

C.

D. Correct: See P&ID M-52-1.

Level: RO Exam


Source: New



Reference(s): M-52-1, “Core Spray P&lD”

HC.OP-AR.ZZ-0007, “Overhead Annunciator Window Box B3” M-51-1, “RHR P&ID”

KA: 209001 .K1.03 Knowledge of the physical connections and/or cause effect relationships between LOW PRESSURE! CORE SPRAY SYSTEM and the following: (CFR 41.2 to 41.9 / 45.7 to 45.8) K1.03 Keep fill system 2.9 3.0



1 Pt(s) The plant is operating at 100% power when CORE SPRAY LINE BREAK (B3-B5) annunciates. This alarm could be the result of:

A. An inadvertent trip of ECCS jockey pump CP228 (RHR A/C and CSS A)

B. An inadvertent SLC injection

C.

D.

Placing the HPCI system in full flow test

Placing Core Spray Loop A in full flow test while Core Spray loop B is idle


A. Incorrect: Sensors detect differential pressure downstream of testable check valves HV-F006NB. With a loss of keep fill “A” loop, the upstream side of HV-FO06NB would see lower pressure but there is no change to downstream conditions. Correct: The alarm response procedure lists this as a cause. SLC injects into the CS loop and the injection flow is sufficient to generate a DP which will activate the sensor that brings in the alarm. Incorrect: HPCI injection would cause this alarm but not full flow test mode. Incorrect: Sensors detect differential prcssure downstream of testable check valves HV-F006A/B. Placing a loop in full flow test would not effect the pressure sensed downstream of HV-F006NB.

Level: RO Exam

B.

C.

D.

Lesson Plan Objective: CSSYSOEOlO From memory, summarizeiidentify the purpose of the Core Spray System line break detection system, IAW thc Core Spray System Lesson Plan.

Source: Modified HC Bank (Q56570) Added a new answer (SLC injection) and this answer is the new correct answer.


Reference(s):


HC.OP-AR.ZZ-0007 for alarm window B3-B5

KA: 211000.K3.02 Knowledge of the effect that a loss or malfunction of the STANDBY LIQUID CONTROL SYSTEM will have on following: (CFR: 41.7 / 45.4) K3.02 Core spray line break detection system: Plant-Specific 3.0* 3.2*




Reactor is critical RPS bus A is being powered from the backup power supply, MCC 10B491 through transformer 1AX432. While restoring RF'S bus A to the normal supply, the RPS Bus Transfer Switch is taken from "ALT A" through "NORMAL" to "ALT B".

SELECT the plant response:

A.

B.

C.

D. FuU reactor scram.

1/2 scram on RPS B, RPS A remains energized.

1/2 scram on RPS A, RPS B remains energized.

RF'S A and RPS B remain energized.


A.

B.

C. D.

Incorrect: RPS "A" would deenergize when placed from ALT A to NORMAL. Incorrect: RPS "B" would deenergize when placed from NORMAL to ALT B. Incorrect: Both channels would be deenergized. Correct: Transferring from ALT A to NORMAL will cause a 112 scram on channel "A" then going to ALT B will cause a 1/2 scram on channel " B . The switch is break before make.

Level: RO Exam

Lesson Plan Objective: RPSOOOE006 Given the appropriate system operating procedure explain the effects on the reactor protection system when the power source is transferred from normal to alternate, and vice versa, IAW the RPS System Operating Procedure.



Reference(s): HC.OP-AB.IC-0003


KA: 212000.K4.04 Knowledge of REACTOR PROTECTION SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) K4.04 The prevention of supplying both RPS buses simultaneously f?om the alternate power source: Plant-Specific 3.1 3.1



1 Pt(s) Complete the following statements:

The IRM detectors operate in the curve. In addition, a decrease in IRM detector argon gas pressure will cause the IRM detectors to be

A. proportional / more

B. proportional /less

C. ionization / more

region of the amplification

sensitive.

D. ionization / less


A.

_________-______-_______I_______________-------------------------------------------

Incorrect: Proportional region is incorrect. In addition, argon is used as a detector ionization gas, so reduced argon gas pressure will yield less ionization events; therefore, the detector is less sensitive. Plausible: The SRMs operate in the proportional region. In addition, argon could be confused for a quench gas, and if it did act as a quench gas, then the detector would be more sensitive with less quench gas. Incorrect: Proportional region is incorrect. Plausible: The SRMs operate in the proportional region. In addition, the second half of the answer is correct. Incorrect: Argon is used as a detector ionization gas, so reduced argon gas pressure will yield less ionization events; therefore, the detector is less sensitive. Plausible: Ionization region is correct. In addition, argon could he confused for a quench gas, and if it did act as a qucnch gas, then the detector would be more sensitive with less quench gas. Correct: IRM detectors operate in the ionization region. Argon is used as a detector ionization gas, so reduced argon gas pressure will yield less ionization events; therefore, the detector is less sensitive.

Level: RO Exam

B.

C.

D.

Lesson Plan Objective: IRMSYSE003 Given a drawing of an IRM Detector, label the simplified drawing of the fission chamber and explain the principles of operation of the detector, IAW Control Room Procedures.


Source: New


Refaence(s): Lesson Plan NOHOIIRMSYS, "IRM System"

KA: 215003.K5.01 Knowledge of the operational implications of the following concepts as they apply to INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : (CFR: 41.5 /45.3) K5.01 Detector operation 2.6 2.7



1 Ptis) Given the following:

The reactor is critical. The Unit is in OPCON 2.

Reactor power is on Range 3 of the IRMs. The Source Range Monitors (SRMs) are being withdrawn from the core. During the SRM withdrawal, *24 VDC SRM power is lost on the A&C SRMs with 0 VDC indicated at 1AD307.

Which one of the following describes the SRM system response?

A. An SRM rod block is generated; A&C SRM detectors stop withdrawing from the core.

An SRM rod block is NOT generated: A&C SRM detectors stop withdrawing from the core.

An SRM rod block is generated; A&C SRM detectors continue to withdraw from the core.

B.

C.

D. An SRM rod block is NOT generated; A&C SRM detectors continue to withdraw from the core.


A. Incorrect: The first half of the answer is correct since the SRM inop trip is active up to IRM Range 8 with the mode switch in startup, so a rod block is generated. However, drive power is from 1OY202 AC power not DC power, so the SRM’s continue to withdraw. Plausible: The first half of the answer is correct. The second half of the answer is a plausible incorrect answe~ and knowledge of the power supplies is needed to select the right answer. Incorrect: 24 VDC supplies the high voltage power supply, so a loss of 24 VDC results in an inop trip. The SRM inop trip is active up to IRM Range 8 with the mode switch in startup, so a rod block is generated. Furthermore, drive power is from 10Y202 AC power not DC power, so the SRM’s continue to withdraw. Plausible: The first half of the answer is plausible and knowledge of the logic and interlocks is required. The second half of the answer is

B.


a plausible incorrect answer and knowledge of the power supplies is needed to select the right answer. Correct: 24 VDC supplies the high voltage power supply, so a loss of 24 VDC results in an inop trip. The SRM inop trip is active up to IRM Range 8 with the mode switch in startup, so a rod block is generated. Drive power is from 10Y202 AC power not DC power, so the SRM’s continue to withdraw. Incorrect: 24 VDC supplies the high voltage power supply, so a loss of 24 VDC results in an inop trip. The SRM inop trip is active up to IRM Range 8 with the mode switch in startup, so a rod block is generated. Plausible: The first half of the answer is a plausible incorrect answer and knowledge of the logic and interlocks is required. The second half of the answer is correct.

C.

D.

Level: RO Exam

Lesson Plan Objective: SRMSYSE013 From memory, predict the SRMS response to loss of operating supply voltages, IAW the Student Handout.

Source: lNP0 Bank (Perry 1)


Reference(s): 0 HC-0P.AB.ZZ-0151, “24 Volt DC Malfunction”

Lesson Plan NOH01 DCELEC, “DC Electrical Distribution” 0 Lesson Plan NOH04SRMSYS, “SRM System”

KA: 215004.K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRH) SYSTEM : (CFR: 41.7 / 45.7) K6.02 24/48 volt D.C. power 3.1 3.3




You are directed to verify the scram and you are aware of industry experiences where several light bulbs on the full core display have failed concurrently.

Which of the following indications would cause you to conclude that the scram may NOT have been successful?

A reactor scram signal has occurred The full core display has several full-in lights NOT displayed CRIDSiSPDSiRWM computer rod position data is NOT available

A.

B.

C.

D.

All LPRM ‘‘downscale’’ lights are illuminated.

Reactor period stable at -80 seconds on SRMs.

RPS channel trip logic lights are extinguished.

All APRM “downscale” lights are extinguished.


A. Incorrect: LPRM downscale lights are not a conclusive indication that the reactor is shutdown; however, all lights lit should are a fairly strong indicator that the reactor is shutdown. Plausible: Incorrect: -80 second period is an indication for a shutdown reactor. Plausible: Incorrect: RPS trip lights indicate that RPS is tripped; however, it does not provide any assurance that the rods actually went in. Plausible: Correct: All APRM downscale lights extinguished means that power is greater than 4% power; therefore, the reactor is NOT shutdown.

B.

C.

D.

Level: RO Exam

Lesson Plan Objective: ABOOOOE003


Source: Hope Creek Bank Last NRC Exam 1999


Reference(s):

KA: 215005.A1.02 Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE MONITOWLOCAL POWER RANGE MONITOR SYSTEM controls including: (CFR: 41.5 / 45.5) A1.02 RPS status 3.9 4.0



The Unit was operating at 100% power. A Station Blackout is in progress. HPCI failed and is unavailable. RCIC is injecting at low flow with RF’V level at + 35 inches. Minimum flow valves SV-F019 & SV-4405 are both open as expected for the low flow condition. The solenoid coil fails on SV-F019. SV-4405 is unaffected.

%ch one of the following describes the effect of the solenoid coil failure on the RCIC system and the appropriate operator action?

A. SV-F019 fails open, so minimum flow protection is maintained. However, the valve will not close if injection flow increases, so operators may need to close the manual block valve downstream of SV-F019 to prevent CST water from being diverted to the suppression pool.

SV-F019 fails closed, so minimum flow protection is lost. The operators must either raise injection flow to maintain adequate pump flow or secure RCIC.

SV-F019 fails open, so minimum flow protection is maintained. No operator action is required since SV-4405 wiU open and close as needed based on injection flow.

SV-F019 fails closed, but minimum flow protection is maintained. No operator action is required since SV-4405 will open and close as needed based on injection flow

B.

C.

D.

__I__--__-________-_I___________________-------------------------------------

Distractcr Analysis:

A. Incorrect: Valve fails closed NOT open. In addition, the second half of the answer implies that SV-F019 and SV-4405 are in a parallel arrangement for min flow redundancy; however, the valves are in series. Plausible: If the valve did actually fail open, and if the valves were in a parallel arrangement, then the answer would be correct. The operator needs to remember the valve arrangement and failure mode.


B. Correct: The valve fails closed. In addition, it is in series with SV- 4405, so minimum flow protection is lost in a situation where it is needed. The operators must either raise injection flow or secure RCIC to prevent pump overheating and damage. Incorrect: Valve fails closed NOT open. Plausible: The second half of the answer is correct. In addition, if the valve did fail open, then the whole answer would be correct. The operator needs to remember the valve failure mode. Incorrect: The answer implies a parallel valve arrangement: however, the valves are actually in a series arrangement, so min flow protection is lost. Plausible: If the valves were in a parallel arrangement, then the answer would be correct. The operator needs to remember the valve arrangement and failure mode.

Level: RO Exam

C.

D.

Lesson Plan Objective: RCICOOE023 Given any of the following and appropriate control room reference material, evaluate and determine the effect on the RCIC system of the following IAW the RCIC System Lesson Plan:

1. A given valve opening or closure

2. Loss of DC or AC power supply

3. Inadequate system flow

4. ETC.

Source: New


Reference( s): Lesson Plan NOH04RCIC00, “RCIC System”

KA: 217000.A2.03 Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR 41.5 / 45.6) A2.03 Valve closures 3.4 3.3




0

The Unit was operating at 100% power. A total Station Blackout is in progress. HPCI failed and is unavailable. RCIC is injecting at low flow with RPV level at + 35 inches. RPV pressure is 120 psig and dropping. HC.OP-AB.ZZ-0135 and the appropriate EOPs are being executed.

Complete the following statement:

RCIC will automatically isolate when RF’V steam supply pressure reaches approximately psig, AND the Emergency Operating Procedures

the isolation to be defeated.

A. 65 ; doNOTallow

B. 65 ; allow

C. 100 ; doNOTallow

D. 100 ; allow

---------------_----I___________________--------------------------


A. Incorrect: EOP-0101 lists RCIC as a preferred injection source and EOP-0101 Table 1 states, “If necessary, bypass the low pressure isolation”. Plausible: Setpoint is correct AND the EOPs do NOT allow the low steam supply pressure isolation to be defeated for HPCI, so it is plausible that someone could conclude the same for RCIC. Correct: The nominal RCIC low steam supply pressure isolation setpoint is approximately 65 psig. EOP-0101 lists RCIC as a preferred injection source and EOP-0101 Table 1 states, “If necessary, bypass the low pressure isolation”. Incorrect: Setpoint is incorrect AND EOP-0101 lists RCIC as a preferred injection source and EOP-0101 Table 1 states, “If necessary, bypass the low pressure isolation”. Plausible: 100 psig is the HPCI low pressure isolation setpoint AND the EOPs do NOT allow the low steam supply pressure

B.

C.


isolation to be defeated for HPCI, so it is plausible that someone could conclude the same for RCIC.

Plausible: 100 psig is the HPCI low pressure isolation setpoint. AND the RCIC isolation may be defeated.

D. Incorrect: Setpoint is incorrect.

Level: RO Exam

Lesson Plan Objective: RCICOOE009 (R) Given plant conditions, determine the signal(s) which will cause the RClC System to automatically isolate, LAW the system operating procedure.

Source: New


Reference(s): Lesson Plan NOH04RCIC00, “RCIC System” HC.OP-ABZZ-0135 HC.OP-SO.BD.OOO1 EOP-0101

KA: 217000.A2.03 Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR. 41.5 / 45.6) A2.03 Valve closures 3.4 3.3



1 Pt(s) The following plant conditions exist at T = 0:

0

Which of the following describes how ADS will respond under these conditions?

Reactor water level is -1 29 inches and slowly dropping Reactor pressure is 900 psig D w e l l pressure is 1.40 psig and steady "A" RHR pump is running All of the MSIVs are closed ADS has NOT been inhibited

A. ADS wiU initiate at T = 405 seconds.

B. ADS will initiate at T = 300 seconds.

C. ADS will initiate at T - 105 seconds.

D. ADS will NOT initiate.


A. Correct: The high drywell pressure bypass timer is activated at -129 inches, so ADS will initiate after the 5 minute (300 second) high drywell pressure bypass timer times out PLUS the 105 second ADS timer. 405 total seconds until initiation. One RHR pump running is sufficient to support ADS initiation. Incorrect: ADS initiates in 405 seconds in this scenario. See justification for answer A. Plausible: 300 seconds is the time for the high drywell pressure bypass timer. This timer is activated in this scenario, but the 105 second ADS timer must also time out after the 300 second bypass timer times out. It is possible that someone could conclude that 300 seconds is the time to initiation. Incorrect: ADS initiates in 405 seconds in this scenario. See justification for answer A. Plausible: -129 inches alone will initiate ADS but only after the 300 second high drywell pressure bypass timer times out followed by the 105 second ADS timer. In addition, if high drywell pressure coincidence were met (1.68 psig), then ADS would initiate in 1 05 seconds.

B.

C.


D. Incorrect: ADS initiates in 405 seconds in this scenario. See justification for answer A. Plausible: A candidate could believe that -129 inches and 1.68 psig are both required for ADS to initiate. In addition, a sufficient number of low pressure ECCS pumps need to be running for ADS to initiate (1 RHR or 2 CS pumps), so someone could think that just 1 RHR pump is not enough.

Level: RO Exam

Lesson Plan Objective: ADSSYSE004 From memory, lisu’identify the five (5) signals (including set points) which will cause the Automatic Depressurization System to automatically initiate, IAW the Automatic Depressurization System Lesson Plan.

Source: HC Bank (456872) NRC exam 12/98 Slightlymodified. Stem was changed to have only 1 RHR pump running rather than all ECCS. Distracter B was changed from “ADS initiates at time 0” to “ADS initiates at T = 300 seconds”.


Reference(s): HC.OP-SO.SN-0001

KA: 218000.A3.04 Ability to monitor automatic operations of the AUTOMATIC DEPRESSURIZATION SYSTEM including: (CFR: 41.7 / 45.7) A3.04 Primary containment pressure 3.7 3.8


9Qoestion: 39 Answer: D

1 Pt(s) The Unit is operating in OPCON 1 when RPV level drops to - 45 inches

Which one of the following describes the expected SPDS RPV Level 2 isolation valve status box display?

A. NORMALinred

B. NORMAL in green

C. ISOLinred

D. ISOL in green


A. Incorrect: There is no indication showing NORMAL in red. NISOL in red indicates that an isolation signal was received hut all of the required valves did not close Plausible: Someone may not recognize that -45 exceeds the level 2 set point, or someone may confuse RPV level 2 with level 1 (-129) inches. Red may he confused with valve position indication for an open valve, so if someone thinks that an isolation was not required, and then they may conclude the valves should be open (red indication). Yet another possibility is that someone could conclude that an isolation was required (red) and that all valves went closed (“normal” for an isolation). Incorrect: NORMAL in green indicates the status without an isolation signal. -45 inches is below the -38 inch RPV level isolation signal. Plausible: Someone may not recognize that -45 exceeds the level 2 set point, or someone may confuse RPV level 2 with level 1 (-129 inches). Incorrect: ISOL is correct, but red would indicate that the safety function did not complete. Plausible: A possible misconception is that ISOL would display in red since an isolation signal implies that something is wrong (red =

danger). Correct: RPV Level 2 corresponds with the - 38 inch isolation signal. SPDS should indicate ISOL in green for RPV Level 2 isolation valves since an isolation signal is in and all valves should he isolated to complete the isolation function. ISOL indicates that an

B.

C.

D.


isolation signal is in and green, in this case, means that valves closed to complete the safety function.

Level: RO Exam

Lesson Plan Objective: NSSSSOE006: Given a copy of the SPDS containment isolation status, evaluate the status of the NSSSS generated isolation signals IAW the NSSSS student handout. NSSSSOE014: Given a specific parameter, which initiates NSSSS isolation signals, identify all valves isolated by that parameter and the set point at which that isolation signal is generated.

Source: New


Reference@): Lesson Plan NOH04SPDSOOC, “SPDS” Lesson Plan NOH04NSSSSOC, Nuclear Steam Supply Shutoff System”

KA: 223002.A4.05 Ability to manually operate andor monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) A4.05 SPDS/ERIS/CRIDS/GDS: Plant- Specific. 2.5* 2.8*




0

0 Reactor power is 97%

Select the IMMEDIATE operator action:

A.

Power ascension is in progress following a refuel outage

PSV-F013P opens inadvertently and does NOT reclose

Depress the "Reset Logic Armed" pushbutton for 'lB" Low- Low set logic.

Lock the mode switch to SHUTDOWN.

Reduce reactor power to 95%.

Dispatch the operator to remove the SRV fuses.

B.

C.

D.


A. Incorrect: Subsequent action. B. Incorrect: Retainment override if unable to close the valve, C. D. Incorrect: Subsequent action.

-__1_---__----------l______l____________------------------------------------

Correct: f i s is an immediate action per HC.OP-AB.WV-0006.

Level: RO Exam

Lesson Plan Objective: ABRF'V6E003 From memory, recall the Immediate Operator Actions for Safetymelief Valve.

Source: HC Bank (477604)


Reference(s): HC.OP-AB.RPV-0006

KA: 239002.G2.4.49 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.101 43.21 45.6)




The Unit is in OPCON 3 following a shutdown for reheling RPV pressure is 900 psig. Feed water level control is maintaining level at + 25 inches. An RPV cool down is in progress.


The feed water level control system uses and indicated RPV level will read range as the RPV cools down.

range instrumentation, than actual RF'V level on that

A. narrow

B. narrow

lower

higher

C. wide I lower

D. wide I higher


A.

B.

........................................................................ -I

Incorrect: Narrow range is correct; however, the second half is incorrect. See justification for answer B. Correct: FWLCS uses the narrow range. If reactor vessel temperature is significantly below calibration conditions, variable leg density and pressure will be higher, causing indicated level to be higher than actual. The Integrated Operating Procedures contain attachments with Vessel Level Instrumentation Temperature Compensation Curves, which allow determination of actual level for off calibration conditions. Incorrect: Wide range is incorrect and indicated level will be higher than actual. See justification for answer B. Incorrect: Wide range is incorrect; however, the second part is correct. See justification for answer B.

C.

D.


Level: RO Exam

Lesson Plan Objective: RXINSTE007 Given any reactor vessel water level range (i.e,, narrow, wide, etc.), determine each of the following:

The calibration criteria (hot or cold) The operational conditions for which the range is intended Any Control Room instrumentation

Source: New


Reference(s): HC.OP-IOZZ-0004, “Shutdown from Rated Power to Cold Shutdown” Lesson Plan NOH04RXINSTC, “Nuclear Boiler Instrumentation”

KA: 259002.K1.09 Knowledge of the physical connections andor cause effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.09 P sat/T sat (compensation) .2.9 3.0




The Unit was operating at 100% power. RCIC is out of service for maintenance. A station blackout event just occurred. RPV level dropped to - 65 inches. HPCI automatically injected to recover RF'V level. RPV level is now stable at + 35 inches. The RBVS & FRVS systems are not running due to loss of power.

Wluch one of the following best describes long term HPCI operations without RBVS & FRVS in service?

A. HPCI will eventually trip on turbine high exhaust pressure as non condensable gases build up in the barometric condenser due to the loss of the barometric condenser vacuum pump discharge flow path. In addition, HPCI operation is undesirable without ventilation in service since airborne contamination levels may eventually rise in the Reactor Building and/or HPCI compartment.

B. HPCI will continue to operate since the barometric condenser vacuum pump discharge flow path will automatically swap to the main condenser due to high backpressure in the HPCI barometric condenser. The swap will preserve HPCI operation and prevent potential airborne contamination in the Reactor Building and/or HPCI compartment.

HPCI will continue to operate since the barometric condenser vacuum pump discharge flow path will automatically swap to the suppression chamber airspace due to high backpressure in the HPCI barometric condenser. The swap will preserve HPCI operation and prevent potential airborne contamination in the Reactor Building and/or HPCI compartment.

HPCI will continue to operate; however, airborne contamination levels may eventually rise in the Reactor Building and/or HPCI compartment as the barometric condenser vacuum pump continues to discharge into the RBVS/FRVS ductwork without ventilation in service.

C.

D.



A. Incorrect: This condition will not cause high turbine exhaust pressure. The barometric condenser services HPCI turbine gland seal leak OK Plausible: The candidate may incorrectly think that the barometric condenser is linked directly with the turbine main process flow. For instance, rhe main turbine discharges to the main condenser, and a loss of main condenser vacuum leads to a main turbine trip. In addition, the second statement in the answer is correct since it is not “desirable” to run HPCI without ventilation in service. Incorrect: HPCI will continue to operate; however, there is no automatic swap feature to realign the vacuum pump discharge back to the main condenser. Plausible: The HPCI vacuum pump normally discharges to the main condenser for test runs, and the discharge flow path automatically realigns to the RBVSRRVS on a HPCl initiation signal. Incorrect: HPCI will continue to operate; however, there is no automatic feature to swap the vacuum pump discharge to the suppression chamber air space. Plausible: The HPCI turbine normally discharges to the suppression pool, and there are vacuum breakers on the discharge line that connect to the suppression chamber air space. In addition, it would almost make sense to discharge a potentially contaminated discharge path back to primary containment when secondary containment treatment systems are not available. Correct: The DC powered HPCI barometric condenser vacuum pump will continue to discharge into the RBVSRRVS ventilation duct. Without ventilation in service, potentially contaminated non condensable gases may eventually leak into the Reactor Building and/or HPCI compartment.

Level: RO Exam

B.

C.

D.

Lesson Plan Objective: HPCIOOEO06: From memory, summarize the interrelationship(s) between the HPCI System and any of the following IAW control room references (Condenser, RBVS, etc.).

Source: New



Reference( s): Lesson Plan NOHOlHPCI00, “High Pressure Coolant Injection” P&ID M-55-1

KA: 261000.K3.04 Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: (CFR: 41.7 /45.6) K3.04 High pressure coolant injection system: Plant-Specific. 3.1 3.1

Note: Hope Creek uses FRVS (Filter Recirculation Ventilation System) rather than SBGTS. This is a plant specific feature at Hope Creek. Fundamentally, the FRVS & SBGTS systems perform a similar function which is to heat the secondary containment atmosphere and provide an elevated release point under accident conditions. The HPCI system interrelationship with FRVS is the same as for plants using SBGTS.




A station blackout has occurred All 500 KV Lines to Hope Creek are de-energized

IAW HCOP-AB.ZZ-0135, Station BlackouUhss of Offsite PowerDiesel Generator Malfunction, which one of the following 500 KV lines is requested first to be re-energized in order to restore power to Hope Creek 13 KV ring bus?

A. Salem 5037 Line

B. New Freedom 5024 Line

C. Red Lion 5015 Line

D. Deans 5021 Line

Distracter Analysis: A.

B. C. D.

---_1-__-__-___1-----------------------------------------------------------------

Correct: AB-I 35 power restoration strategy is to restore power via the Salem 5037 line and the Salem Gas Turbine. Incorrect: See justification for A. Incorrect: See justification for A. Incorrect: See justification for A.

Level: RO Exam

Lesson Plan Objective: OAB135E006 Explain the information contained in the Discussion Section of Station Blackouthss Of Offsite Power Diesel Generator Malfunction, Abnormal Operating Procedure.

Source: HC Bank (Q68928) NRC Exam 2/2002


Reference(s): HC.OP-AB.ZZ-0135,4.14

KA: 226001.K2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) K2.01 Off-site sources ofpower 3.3 3.6




120 VAC unintermptible power supply 1AD481 is aligned with the MANUAL BYPASS CONTROL SWITCH in “NORMAL”. A loss of voltage is detected on the output of the inverter section.


The static switch will automatically transfer to the supply, and the static switch inverter once the inverter output voltage returns to normal.

power automatically swap back to the

A. alternate /will

B. alternate / will not

C. backup/will

D. backup /will not


A. Incorrect: The altemate source is an auctioneered supply on the input side of the inverter section. In this scenario, the inverter output section went dead, so both the normal and altemate supplies are not available. Thmefore. the static switch will automatically swap to the backup source in this particular case. The second half is correct in that the static switch will swap to the “preferred source” (inverter output) once voltage retums to normal. Incorrect: The alternate source is an auctioneered supply on the input side of the inverter section. In ths scenario, the inverter output section went dead, so both the normal and alternate supplies are not available. Therefore, the static switch will automatically swap to the backup source in this particular case. The second half is also incorrect since the static switch will swap to the “preferred source” (inverter output) once voltage returns to normal. Correct: In ths scenario, the inverter output section went dead, so both the normal and altemate supplies are not available. Therefore, the static switch will automatically swap to the backup source in this particular case. In addition. the static switch is normal seeking and will swap to the “preferred source’’ (inverter output) once voltage retums to normal.

B.

C.


D. Incorrect: The first half is correct, but the second half is incorrect. See justification for answer C.

Level: RO Exam

Lesson Plan Objective: 1EACOOE020 Given a set of plant conditions, summarizelidentify the UPS System response to the following:

Loss of AC to the Rectifier Cabinet Loss of DC to the Rectifier Cabinet

Loss of AC Regulator Output Loss of Inverter output

Source: New


Reference(s): Lesson Plan NOHOIEACOO, “Class 1E AC Power Distribution”

KA: 226001.K4.01 Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C.D.C.) design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) K4.01 Transfer from preferred power to alternate power supplies 3.1 3.4




The Unit is in OPCON 1 at 100% power. The CONTROL AREA HVAC FAN MALF (E6-BI) is in alarm. A & BVH410, Control Area Battery Exhaust Fans have both failed and will be out of service for the next 24 hours. Hydrogen monitoring is being performed in the affected battery rooms once per shift in accordance with HC.OB-AELHVAC-0001, "HVAC". During rounds, the responsible operator reports that the current HPCI battery room Hydrogen concentration is about I%, and this H2 concentration is equivalent to 25% of the Lower Explosive Limit (LEL).

Which one of the following describes the correct course of action?

A. Initiate a fire impairment and prop open the doors to the HPCI battery room to reduce the Hydrogen concentration.

B.

C.

Secure the HPCI battery charger to stop H2 generation.

Place the HPCI battery charger in EQUALIZE mode to minimize H2 generation.

Initiate a fire impairment to maintain the HPCI battery room doors closed and to restrict access to the room until the exhaust fans are restored.

D.


A. Correct: This is the only option presented that both reduces the H2 concentration in the room while preserving battery and HPCI function. In addition, procedure HC.OB-AB.HVAC-0001, "HVAC" specifically directs that the doors be opened if H2 is above 10% of the LEL. The procedure also directs that a fire impairment be initiated due to the need to prop open the doors. . Incorrect: This not a course of action directed by any procedure. This action would render the battery and HPCI inoperable (see Tech Spec 3.8.2.1 Action b.). lt would also affect availabilityduring any event requiring HPCI since the batteries would discharge.

B.


Furthermore, the action does nothing to correct the existing high H2 concentration. Plausible: Battery charging does result in H2 generation, so securing the charger may seem to be a possible solution. Incorrect: Equalize mode places an overcharge condition on the battery; therefore, the H2 generation rate and concentration will increase. Plausible: It is a mode of charger operation and could be selected if this mode of operation and its consequences were not understood. Incorrect: Maintaining the current situation will allow the H2 concentration to increase to potentially flammable conditions. Plausible: Containing the situation to the one room may seem like a good idea. Restricting access is a good idea in this situation.

Level: RO Exam

Lesson Plan Objective: DCELECE018 Given a set of plant conditiondmalfunctions associated with battery ventilation, evaluate whether a loss of ventilation to a battery room can result in equipment failure, L4W Available Control Room References.

C.

D.

Source: New


Reference(s): HC.OP-AB.HVAC-0001, “HVAC” Lesson Plan NOH01 CAVENTC, “Control Area Ventilation” Lesson Plan NOHOlDCELEC, “DC Electrical Distribution”

KA: 263000.K5.01 Knowledge of the operational implications of the following concepts as they apply to D.C. ELECTRICAL DISTRIBUTION: (CFR: 41 .S / 45.3) K5.01 Hydrogen generation during battery charging. 2.6 2.9




0

0

EDG-A is paralleled to 10A401 Bus for testing. The engine driven fuel oil pump fails due to pump seizure.

What will be the EDG frequency and diesel response to this condition?

EDG output frequency will., . .

A.

B.

C.

remain constant UNTIL the diesel trips on reverse power.

lower AND THEN the diesel will trip on low fuel oil pressure.

lower AND THEN the diesel will trip on reverse power.

D. remain constant AND the diesel will continue to run.

-_--_____---_--__--________________I____--------------------------------------------


A. Incorrect: The electric standby fuel oil pump will automatically start when fuel oil pressure drops to 20 psig, so the engine will continue to run as if nothing happened. Plausible: If there were no standby fuel oil pump, then this would be the correct answer. Frequency would remain constant, the diesel would stop carrying load, and the generator would motor until it tripped on reverse power. Incorrect: The electric standby fuel oil pump will automatically start when fuel oil pressure drops to 20 psig, so the engine will continue to run as if nothing happened. Plausible: A plausible misconception is that frequency would lower as fuel starvation occurred. In addition, some engine designs have a low fuel oil pressure trip, but this one does not. It does have a low lube oil pressure trip. So, the second half of the answer is plausible but incorrect. Incorrect: The electric standby fuel oil pump will automatically start when fuel oil pressure drops to 20 psig, so the engine will continue to run as if nothing happened. Plausible: A plausible misconception is that fr'equency would lower as fuel starvation occurred. In addition, if there were no standby pump, then the diesel would in fact motor and trip on reverse power.

B.

C .


D. Correct: Frequency is locked in a grid frequency when the EDG is paralleled to the grid. More importantly, the electric standby fuel oil pump will automatically start when fuel oil pressure drops to 20 psig, so the engine will continue to run as if nothing happened.

Level: RO Exam

Lesson Plan Objective: EDGOOOE007 Given plant conditions, determine if the Diesel Generator will trip under manual and/or automatic start conditions.

Source: Mod HC Bank (Q25853) Stem condition was modified to change the malfunction from a governor problem to a fuel oil pump failure. Changed the answer to be that the engine would continue to run. Replaced one distracter. The new distracter states that the diesel will trip on low fuel oil pressure rather than generator overcurrent.


Reference(s): Lesson Plan NOH0 1 EDGOOOC, “Emergency Diesel Generators”

KA: 264000.K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the EMERGENCY GENERATORS (DIESELLTET) : (CFR: 41.7 / 45.7) K6.02 Fuel oil pumps 3.6 3.6




0

The Unit is in OPCON 1. Instrument Air Dryer 00-F-104 is IN SERVICE. Instrument Air Dryer 10-F-104 is in STANDBY. HV-I 1416, “Instrument Air Dryer 1A-F-104 Isolation Valve” is CLOSED with hand switch HS-I 1416 in CLOSE. Instrument Air header pressure is at about 100 psig. FSH-7618 fails high and generates an EXCESSIVE FLOW signal. 0

Which one of the following describes the expected initial Instrument Air system response, and if required, any appropriate operator actions that may be needed to stabilize or restore Instrument Air header pressure?

A. Instrument Air header pressure will slowly drop. Place HS- 11416 in OPEN for HV-11416 “Instrument Air Dryer 1A-F-104 Isolation Valve”.

B. Instrument Air header pressure will remain in the normal range since HV-11416 will automatically OPEN immediately upon receipt of the EXCESSIVE FLOW signal. No operator action is required.

C. Instrument Air header pressure will slowly drop and HV-11416 will automatically OPEN when Instrument Air header pressure drops to 90 psig. No operator action is required.

Instrument Air header pressure will slowly drop. JACK OPEN HV-11416 by isolating the vent off the bottom of the air piston since the valve will NOT open either automatically OR with the control switch as a result of the EXCESSIVE FLOW signal.

D.


A. Correct: The false EXCESSIVE FLOW signal will trip off the in service dryer (00-F-104) and will prevent the automatic swap to the standby filter (10-F-104). Air header pressure will slowly decay due to normal air usage. Since there is no leak, then the operators can place the third air dryer (1 A-F- 104) by opening the associated discharge valve (HV-I 1416) via the hand switch. This will stabilize and restore header pressure. Reference HC.OP.AB.COMP.


B. Incorrect: There is no automatic open feature based on EXCESSIVE FLOW. Plausible: The correct action is to open HV-I 1416 under these circumstances, and there are some dryer malfunctions that will cause the valve to automatically open. Incorrect: HV-I 1416 will not automatically open at 90 psig. Plausible: Air header pressure will slowly drop and the valve will open automatically at 85 psig. Incorrect: HV-I 1416 fails open on loss of air, so this action would actually force the valve to close rather than open. Plausible: Instrument air header pressure will drop. The action would be correct to attempt to open the discharge valve for the dryer that was in service (00-F-104) or for the dryer that was in standby (10-F-104) as directed by HC.OP.AB.COMP.

Level: RO Exam

C.

D.

Lesson Plan Objective: INSAlREOl3 From memory, describe the Instrument Air System dryer package response if the in-service dryer package were to receive an AUTO TRANSFER signal when any of the following conditions existed, available references.

0

Standby dryer package has an alarm condition. Operating in-service dryer package has air excessive flow condition. Off-service dryer package control switch is not in the STANDBY position.

Source: New


Reference( s): 0

NOHOIINSAIR, ‘‘Instrument Air System”

HC.OP-AB.COMP-0001, “Instrument Air and/or Service Air”

KA: 300000.A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: (CFR: 41.5 / 45.6) A2.01 Air dryer and filter malfunctions . 2.9 2.8




The plant is at 86% power. ‘A’ SACS Loop is supplying TACS. ‘D’ SACS pump is operating in the ‘B’ SACS Loop. ‘B’ SACS pump is in AUTO. ‘A’, ‘C’ and ‘D’ Service Water pumps are running. A guillotine rupture of the TACS piping occurs causing a Low-Low pressure in the supply accumulator.

Which of the following is the system response?

A. The ‘B’ SACS pump will start, the ‘B’ loop SACS to TACS isolation valves will open, leaving the ‘A’ and ‘B’ SACS loops cross-connected.

B. The ‘A’ and ‘C’ SACS pumps will trip on low flow, the ‘B’ SACS pump will start, and the ‘B’ pump SACS to TACS isolation valve will open.

C. The ‘B’ SACS pump will start, the ‘B’ and ‘D’ SACS to TACS isolation valves will open, and the ‘A’ loop SACS to TACS isolation valves will shut.

D. All SACS pumps will trip on low SACS pump flow.


A. Correct: HC.OP-SO.EG-0001 interlock section, closure of the E/F will cause a low loop flow which will stat the auto pump and open the opposite loop valves. Result is the loops are cross-connected. Plausible: Incorrect: Pumps trip on high flow, not low flow. Plausible: Incorrect: “A” through “D’ valves need a pump trip or isolation signal to close. Plausible: Incorrect: Pumps trip on high flow, not low flow. Plausible:

B.

C.

D.


Level: RO Exam

Lesson Plan Objective: STACSOEOl5 Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the Main Control Room; assess the status of the STACS or its components by evaluation of the controls/instrumentation/alarms. IAW available control room references

Source: HC Bank 45623 1


Reference(s): HC.OP-SO.EG-0001

KA: 400000.A1.03 Ability to predict and / or monitor changes in parameters associated with operating the CCWS controls including: (CFR: 41.5 / 45.5) A1.03 CCW Pressure 2.7 2.7



1 Pt(s) Following a full Group I isolation at full power, HPCI received an auto initiation signal on RPV low level. During the HPCI initiation on low level, the HPCI pump flow transmitter output signal failed and was sending a constant ZERO gpm flow signal to the controller.

Which of the following describes the HPCI system response if NO operator action is taken?

The HPCI turbine will ...

A.

B.

trip on overspeed and remain shutdown.

trip as soon as the signal is lost.

C.

D. operate at minimum speed.

trip on high RPV water level.


A. Incorrect: May trip on mechanical overspeed depending on pump load but restarts after OS automatically resets. Plausible: Incorrect: No trip signal from loss of flow. Plausible: Correct: With the flow controller output sensing zero flow, the output will be maximum speed demand. HPCI will develop full flow until +54 inch high level trip is reached. Incorrect: If the flow is sensed low, the governor will try to speed up to establish 5600 gpm. Plausible:

B.

C.

D.

Level: RO Exam

Lesson Plan Objective: HPCIOOEO16 (R) Given plant conditions and a drawing of the controls, instrumentation andor alarms located in the main control room, assess the status of the HPCI System by evaluation of the controls/instrumentation/alarms, IAW control room references.




Reference(s): M-56, HC.OP-SO.BJ-0001

KA: 206000.A3.02 Ability to monitor automatic operations of the HIGH PRESSURE COOLANT INJECTION SYSTEM including: (CFR: 41.7 / 45.7) A3.02 System Flow: BWR-2,3,4 3.8 3.8



Which of the following correctly describes the proper method for resetting an overspeed trip of the RCIC turbine?

A.

1 Pt(s)

The trip valve actuator is closed from the control room and the overspeed trip linkage must he manually relatched locally at the actuator and valve stem. The valve is then reopened from the control room.

B. The trip valve actuator is closed locally and the overspeed trip linkage must be manually relatched locally at the actuator and valve stem. The valve is then reopened locally.

C. The trip valve actuator is closed locally and the valve relatches automatically. The valve is then reopened locally.

The trip valve actuator is closed from the control room and the valve relatches automatically. The valve is then reopened from the control room.

D.


A. Correct: The trip valve actuator is closed from the control room and the overspeed trip linkage must be manually relatched locally at the actuator and valve stem. The valve is then reopened from the control room. Reference HC.OP-SO.BD-001 Sect 5.7 Incorrect: See justification for A. Incorrect: See justification for A. Incorrect: See justification for A.

Level: RO Exam

B. C. D.

Lesson Plan Objective: RCICOOE005 Given plant conditions regarding the RCIC trip and throttle valve:

rn

0

Distinguish a mechanical overspeed trip from other trips IAW the RCIC System Lesson Plan. Explain how to reset a mechanical overspeed trip IAW the system operating procedure. Explain how the valve closes on a trip signal IAW the RCIC System Lesson Plan.




Reference(s): HC.OP-SO.BD-001 Sect 5.7

KA. 217000.A4.02 Ability to manually operate andor monitor in the control room: (CFR: 41.7 / 45.5 to 45.8) A4.02 Turbine trip throttle valve reset 3.9 3.9



1 Pt(s) Which one of the following describes the purpose of the Shutdown Cooling Return Check Valves (HV-FOSOA/B)?

A. Prevents bypassing SDC flow around the core through the idle SDC loop so that the RPV does not inadvertently heat up.

Limits RPV inventory loss in the event that the RHR pump suction valve (F004) is inadvertently opened for the pump running in the SDC loop.

Limits RPV inventory loss in the event of an RHR line break on the FO15A(B) side of the check valve.

Ensures that all SDC flow is redirected to the LPCI injection line on a LPCI initiation since the SDC Return Check Valve will “check” closed when the LPCI injection valve (FO17) opens in the loop running in SDC.

B.

C.

D.


A. Incorrect: SDC core bypass is a concern through the Recirc Pump discharge and suction valves. In addition, the F015 motor operated valve is isolated on the idle loop and that valve is in series with the check valve, so there is no potential short circuit flow path. Plausible: SDC bypass flow is a significant concern during SDC ops, and Hope Creek has Operating Experience with this problem. In addition, P&ID’s and simplified drawings make it appear as though this scenario were plausible. Incorrect: The SDC Return Check Valve would not mitigate this event since the flow path from the RPV to the Suppression Pool is on the suction side of the pump in this particular scenario. Also, with the pump running, flow will continue into the RPV through the SDC retum line. Plausible: The purpose of the check valve is to limit RPV inventory loss and there would be aphysical path from the Suppression Pool to the RPV via the SDC retum line; however, the drain down flow path is from the SDC SUCTION line (from the Recirc Loop) to the SP via the F008, F009, F006, and F004. Correct: The valve will check closed in the event of a line break to limit RPV inventory loss. Refer to the lesson plan.

B.

C .


D. Incorrect: The RHR loop does not fully realign from SDC mode to LPCI mode. The F004 suction valves are closed and do not get an open signal. On a low level LPCI injection, the pumps wind up tripping on loss of suction path. Also, the check valve does not "check" in the LPCI flow direction. Plausible: RHR does automatically realign from test mode to injection mode. In addition, if RHR did align from SDC to LPCI, then it would be important to redirect flow from SDC to LPCl (the motor operated SDC return valve does in fact close on a LPCI signal).

Level: RO Exam

Lesson Plan Objective: RHRSYSE003 Given a copy of P&ID M-51-1, determine each of the following operating modes of the RHR System: a. Low Pressure Coolant Injection b. Suppression Pool CoolingiFull Flow Test c. Drywell Spray d. Suppression Chamber Spray e. Fuel Pool Cooling Assist f. Shutdown Cooling g. Normal Standby Alignment Source: New


Reference( s): Lesson Plan NOHOlRHRSYSC, "RHR'

KA: 205000.G2.1.28 2.1.28 Knowledge ofthe purpose and function of major system components and controls. (CFR: 41.7)


Question: 52 Answer: 6

1 Pt(s) 125 VDC Bus B (1BD417) tripped off due to a bus fault.

From which location(s) can SRV F “Foxtrot” be opened if needed?

A. Available from the Main Control Room and Remote Shutdown Panel.

Not available from any location.

Available Main Control Room only.

Available from the Remote Shutdown Panel only.

B.

C.

D.

-----_-----_-------_______l____l________-----------------------------------


A. Incorrect: Cannot he opened from either location. See justification for answer B. Plausible: ADS SRV’s (A,B,C,D,E) each have an “A” coil and a “ B coil, so if F were mistaken for an ADS SRV, then someone might conclude that power was available from lDD417. In addition, the valve does have control stations at both the MCR and RSP.

non-ADS SRV, so it only has one solenoid and it is an “A” solenoid. SRV F has one control station in the MCR and one at the RSP; however, the valve cannot be opened from either location since both locations use the same “A” solenoid and there is no power available to this SRV solenoid. Incorrect: Cannot be opened kom the MCR. See justification for answer B. Plausible: ADS SRV’s (A,B,C,D,E) each have an “A” coil and a “B’ coil, so if F were mistaken for an ADS SRV, then someone might conclude that power was available from 1DD417. Furthermore, all ADS SRV solenoid valves have 2 control stations in the MCR one for the “A” solenoid and one for the “B” solenoid. Incorrect: Cannot be opened from the RSP. See justification for answer B. Plausible: There is a control station at the RSP for this valve. It is possible that someone could conclude that there is a “B’ solenoid that could be energized via the RSP panel. Please note that ADS valves A & E does have a control station in the lower relay room that

B. Correct: 1BD417 feeds all ofthe SRV “A” solenoids. SRV F is a

C.

D.


uses a "B' solenoid powered from 1DD217, so someone could think that the F SRV might have the same setup at the RSP.

Level: RO Exam

Lesson Plan Objective: MSTEAME003 Concerning the safety relief valves; summarize, list or identify the following.

0

0

The number and type of SRV's at Hope Creek. Which SRV's have an ADS function. Power supplies to the SRV solenoids. Which SRV's can be operated remotely and the location from which each of these valves can be operated. Purpose of the low-low set function and determine which SRV's are used for this function. Determine the sequence of operation of the low-low set SRV's including arming setpoints, lift points and reclose setpoints.

Source: Modified INPO Bank question (LGS)


ReferencHs): Lesson Plan

KA: 263000.K2.01 Knowledge of electrical power supplies to the following: (CFR: 41.7) K2.01 Major D.C. loads. 3.1 3.4



Unintermptible Power Supply 1BD483 fails resulting in a complete loss of power to 1BJ483 distribution panel.

Which one of the following describes the feedwater control system response?

A.

B.

1 Pt(s)

ALL RFPs trip due to the power interruption to DFCS.

All RFPs remain running since DFCS remains in service on the secondary power supply.

“B” RFP trips on loss of speed signal.

“B” RFP trips on loss of control power.

C.

D.


A. Incorrect: DFCS will remain in service on secondary DC power. This is an unintermptible source, so there is neither a short or prolonged interruption in power. The A & C pumps remain running. Plausible: If there was not a secondary power supply or if there was an interruption / loss of power to DFCS, then all the RFPs would trip. Incorrect: “B’ RFP will trip on loss of speed signal. Plausible: DFCS does remain in service on secondary DC power. Correct: A loss of AC distribution panel will result in a loss of speed pickup signal on “B’ RFP, and this results in a RFP trip. DFCS (Digital FW Control System) has a secondary power supply from the DC system. so DFCS will control the 2 remaining RFPs. Incorrect: The loss of power trip is related to a loss of DC control power NOT AC power. Plausible: The trip is similar to the loss of speed signal trip.

Level: RO Exam

B.

C.

D.


Lesson Plan Objective: FWCONTE016 Given any of the following systems, state the interrelationship between the FWLC System and that System, IAW the Feedwater Control System Lesson Plan:

Main Turbine Recirculation System Rod Worth Minimizer (RWM) Main Steam Redundant Reactivity Control (RRCS)

120 VAC Non-1E Electrical Distribution 125 VDC Non-1E Electrical Distribution

Hydrogen Water Chemical Injection System

Source: New


Reference(s): Lesson Plan NOH04FWCONTC, “Feedwater Control System”

KA: 262002.K1.01 Knowledge of the physical connections and/or cause effect relationships between UNINTERRUPTABLE POWER SUPPLY (A.C.D.C.) and the following: K1.O1 Feedwater level control: Plant-Specific. 2.8 3.0



1 Pt(s) Two minutes aAer the reactor scrammed at 12.5" due to a loss oftwo reactor feed pumps, the RO notes the CRD system flow control valve is closed but CRD system flow indication is off scale high.

Under these conditions, the RO should:

A. Inform the CRS and have an N E 0 investigate the faulty valve position indication.

B. Trip the operating CRD pump as soon as all control rods are verified to be fully inserted to prevent pump runout.

Take manual control of the flow control valve and reduce CRD system flow to 65 gpm.

Continue with scram actions since the CRD indications are normal.

C.

D.

______I___________-_----------------------------------------------------------------


A.

B.

Incorrect: Condition is normal. There is no need to investigate. Plausible: Incorrect: There is no need to trip the pump. The system's orifice is designed to prevent runout during a scram condition. Plausible: Incorrect: Condition is normal. There is no need to take manual control. Plausible: Correct: Accumulator flow will cause indication to be high, until the SCRAM is reset.

C.

D.


Level: RO Exam

Lesson Plan Objective: CRMECHE002 (R) Given a simplified diagram of a CRDM, explain the flowpath of water through the mechanism during the following modes of operation:

1. Insert 2. Withdrawal 3. Scram

a. With accumulator water b. With reactor water

4. No Rod Motion (Cooling)



Reference(s): NOH05000006C - Control Rod Drive Hydraulics

KA: 201001.K5.02 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD DRIVE HYDRAULIC SYSTEM: (CFR: 41.5 / 45.3) K5.02 Flow Indication 2.6 2.6.



1 Pt(s) The Unit is at 100% power with the CRD Hydraulic System lined up for normal operations. Reactor pressure is lowered, and it is the onlyparameter that is adjusted.


IF a control rod is given a normal insert or withdrawal command, THEN the insert speed will be than they were before reactor pressure was lowered.

A. slower / faster

and the withdrawal speed will be

B. faster / slower

C. slower / slower

D. faster / faster


A. Incorrect: The only thing that is significantly different between an insert and withdrawal command is that the dnve piston is subjected to drive water pressure on the bottom of the piston, and the top side of the piston is exposed to the exhaust header (reactor pressure) to insert the rod. The reverse is true to withdraw the rod. So, this scenario results in a higher dp across the drive piston, so the speeds increase in both directions. Incorrect: Speeds increase in both directions. (See justification for answer A). Incorrect: The reduction in reactor pressure will result in a hgher dp across the drive piston (see discussion for answer D). Therefore, insert and withdrawal speeds will be faster NOT slower. Correct: Control rod normal insert & withdrawal speeds will be faster after the reactor pressure reduction due to a higher delta p across the drive piston. The pressure control valve is set by throttling the pressure control motor operated valve to the desired position, and the PCV does not automatically modulate to regulate the delta p. In addition, the insert and withdrawal speeds were set via needle valves that were set under previously established system conditions. In short, a reduction in reactor pressure alone will result

B.

C.

D.


in a hgher delta p between drive pressure and reactor pressure (exhaust header), so both insert & withdrawal speeds will be faster.

Level: RO Exam


Source: New


Reference(s): NOH04CRDHYD, “Control Rod Hydraulics”

KA: 2001003.K6.02 Knowledge ofthe effect that a loss or malfunction of the following will have on the CONTROL ROD AND DRIVE MECHANISM : (CFR: 41.7 145.7) K6.02 Reactor pressure 3.0 3.0



1 Pt(s) Recirculation loop flow mismatch limits are based on "effective core flow."

Effective core flow is determined by using:

A. total jet pump flow.

B.

C.

the higher recirc loop flow and multiplying by 2.

the lower recirc loop flow and multiplying by 2.

D.


A. Incorrect: Plausible:

B. Incorrect: Plausible:

C. Correct: See HC.OP-ST.BB-0001.

the average of the two recirc loop flows. ____________________________I___________----------------------------------------

"Recirculation Loop Flow" refers to Jet Pump Loop Flow, not Recirculation Pump Flow. TIS define effective core flow as the Total Core Flow that would result if both Jet pump Loop Flows were assumed to be at the smaller value of the two flows. Example. - If Jet Pump Loop "A" Flow is 48 Mlbmihr and Jet Pump Loop "B" Flow is 46 Mlbmihr, the effective core flow would be 92 Mlbmihr (46 x 2).

D. Incorrect: Plausible:

Level: RO Exam

Lesson Plan Objective: IOP006E003 (R) Apply Precautions, Limitations and Notes while executing the POWER CHANGES DURING OPERATION Integrated Operating Procedure.




Reference(s): HC.OP-ST.BB-OOOl(Q) - Rev. 36 Step 5.1.4.F note

KA: 202002.A1.07 Ability to predict andor monitor changes in parameters associated with operating the RECIRCULATION FLOW CONTROL SYSTEM controls including: (CFR: 41.5 / 45.5) A1.07 Recirculation loop flow: Plant-Specific 3.1 3.1



A reactor startup is in progress with the reactor above the point of adding heat. RWCU was placed in blowdown operation to support level control with HV-F033 HIC-R606 DRAIN FLOW CONTROL throttled about 50% open, The operator notices that the NRHX outlet temperature is 128'F and rising.

What is the correct action to take to reduce the NRHX outlet temperature IAW HC-OP-SO.BG-0001, and what is one operational concern regarding the hgh temperature condition?

Throttle HV-F033 HIC-R606 in the direction. will automatically isolate if the NRHX outlet temperature reaches 140'F.

A. Closed ; HV-F004, Outboard Isolation Valve

B. Open ; HV-F001, Inboard Isolation Valve

C. Closed ; HV-FOO1, Inboard Isolation Valve

1 Pt(s)

D. Open ; HV-FOO4, Outboard Isolation Valve


A. Correct: IAW HC-OF'-SO.BG-0001 B. Incorrect: Opening the valve will increase blowdown flow along

with the heat load on the NRHX causing NRHX outlet temperature to rise. In addition, HV-F001 will NOT isolate on high NRHX outlet temperature. Plausible: Incorrect: The first half is correct; however, HV-F001 will NOT isolate on high NRHX outlet temperature. Plausible: Incorrect: Opening the valve will increase blowdown flow along with the heat load on the NRHX causing NRHX outlet temperature to rise. The second half is correct. Plausible:

C.

D.


Level: RO Exam

Lesson Plan Objective: RWCUOOE018 (R) Given a set of plant conditions evaluate the effects of RWCU System blowdown operation on the RHX and NRHXs IAW the RWCU System Lesson Plan.

Source: New


Reference(s): HC-OP-SO.BG-0001 rev. 40 Section 3.3.1 & Step 5.4.8 HC.OP-1O.ZZ-003 rev. 76 Section 3.3.5

KA: 204000.K2.14 Ability to (a) predict the impacts of the following on the REACTOR WATER CLEANUP SYSTEM ; and (b) based on those predictions, use procedures to correct. control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) A2.14 System high temperature 3.2 3.2




OPCON 1 at 100% power. All control rods are fully withdrawn. The Reactor Operator begins HC.OP-STSF-003, “RPS Manual SCRAM Test - Weekly”. The Reactor Operator arms and presses the RPS TRIP S Y S A1 MAN INIT PB. Coincidentally, the PILOT SCRAM VALVE TRIP ACTUATOR LOGIC Group 4B SOLENOIDS LOGIC B NORMAL light extinguishes at 10C65 tC. This was NOT caused by the previous Operator action.

Which of the following best describes the expected indications on the full core display?

A. Control Rod Groups 1,2, & 3 RED lamps are lit, and Control Rod Group 4 GREEN, BLUE, & AMBER lamps are lit.

All Control Rods have RED lamps lit.

All Control Rods have GREEN, BLUE, & AMBER lamps lit.

Control Rod Groups 1 & 4 GREEN, BLUE, & AMBER lamps are lit, and Control Rod Groups 2 & 3 have RED lamps lit.

B.

C.

D.


A. Correct: The Operator’s action causes an “A” side half scram and all “A” side scram pilot solenoids de-energize. No rod motion occurs. Now, the Group 4B solenoid logic status light extinguishes indicating a loss of power to “B’ side scram pilot solenoids for all control rods in Group 4 only. So, the “A” and “B” scram pilot solenoid valves are both de-energized on all control rods in Group 4, and this vents air from all scram valves for control rods in Group 4 ONLY. The Full Core Display will show RED (full out indication) for rods in Groups 1,2, & 3, and Group 4 control rods will display GREEN (full in), BLUE (scram valves open), and AMBER (rod drift). Incorrect: RED indicates h l l out position. Group 4 control rods went full in, so this answer is incorrect.

B.


Plausible: One could conclude that sufficient logic did not complete to cause any of the rods to scram. Incorrect: This would be an indication of a full scram; however, this problem only caused Group 4 rods to scram in. Plausible: It is possible that someone could conclude sufficient logic made up to cause a full core scram. Incorrect: As stated above, scram logic de-energized to the extent that ONLY the Group 4 control rods scrammed. Plausible: The nomenclature could lead a candidate to conclude that the control rods associated with Groups 1 & 4 scrammed (A1 manual pushbutton was depressed and the Group 4B logic power failure, i.e. WS A & B de-energized on Groups 1 & 4). The justification for the correct answer describes how the logic actually works.

C.

D.

Level: RO Exam

Lesson Plan Objective: MANCONE002

Source: New


Refkrence(s): Lesson Plan NOHOIRPSOOC, “Reactor Protection System” Lesson Plan NOH04MANCONC, “Reactor Manual Control

HC.OP-STSF-0003, “RPS Manual Scram Test - Weekly”

KA: 214000.A3.01 Ability to monitor automatic operations of the ROD POSITION INFORMATION SYSTEM including: (CFR: 41.7 / 45.7) A3.01 Full core display 3.4 3.3

System”




0

A LOCA has occurred in the Drywell W V level is +20 inches and is steady Suppression Chamber pressure is 7.3 psig and rising slowly The CRS directs "B" RHR be placed in Suppression Chamber Spray

Which of the following operator actions and plant conditions are required to establish flow through the Suppression Chamber Spray valve(s) BC-HV- F027B?

The Suppression Chamber Spray valve@) BC-HV-F027B Auto ( 1 ) Overide must be pressed and the (2)

must be (3 )

A. (1) Open, (2) LPCI initiation signal, (3) reset

B. (1) Close, (2) LPCI initiation signal, (3) reset

C. (1) Close, (2) High Drywell pressure signal, (3) present

D. (1) Open, (2) High Drywell pressure signal, (3) present ~~


A.

B. Incorrect: Items 2 & 3 are incorrect.

C. Correct D. Incorrect: Item 1 is incorrect.

Incorrect: All three selections are incorrect. Plausible:

Plausible:

Plausible:

Level: RO Exam

Lesson Plan Objective: FWRSYSE014

Source: HC Bank 456222 (Audit 1999)


Reference(s): HC.OP-SO.BC-0001


KA: 230000 Ability to manually operate and/or monitor in the control room: (CFR: 41.7 145.5 to 45.8) A4.02 Spray Valves 3.8 3.6


Question: 60 Answer: D ~~ ~ -~

1 Pt(s) The unit is shutdown in OPCON 4.

SELECT the system capable of providing emergency makeup to the spent fuel pool that can be completely aligned from the control room.

A. Safety Auxiliary Cooling System.

B. Fire protection.

C. Residual Heat Removal.

D. Service Water System. ~~ ~~~ ~~


A.

B.

C.

Incorrect: Not able to supply emergency fill. Plausible: Incorrect: Cannot be aligned for the control room. Plausible: Incorrect: Not available until Refueling operations via LPCI from Suppression Pool or manual valve operation from the field. Plausible: Correct: Can be aligned via valves form the control room.

Level: RO Exam

Lesson Plan Objective: FPCCOOEOOS



Reference(s): M-10-1 and M-53-1

KA: 233000.G2.1.30 Ability to locate and operate components / including local controls. (CFR: 41.7 / 45.7) 3.9 3.4.

D.



Which of the following conditions will result in an automatic closure of the Main Steam Supply Valve (AB-HV-2016AB) to the steam jet air ejectors?

A. Low offgas flow to the third stage air ejector for thirty seconds and the first stage air ejector suction valves (CGHV- 1968Al,AZ,A3/Bl,BZ,B3) are open.

Low offgas flow to the third stage air ejector for thirty seconds and the third stage air ejector suction valve (CGHV-l956A!B) is open.

Low main steam flow to the third stage air ejector for thirty seconds and the first stage air ejector suction valves (CGHV- 1968Al,AZfi3/Bl,BZ,B3) are open.

Low main steam flow to the third stage air ejector for thirty seconds and the third stage air ejector suction valve (CGHV- 1956AB) is open.

1 Pt(s)

B.

C.

D.


A.

B.

C. Correct: IAW HC.OP-SO.CG-0001. D.

Incorrect: Low offgas flow is incorrect. Plausible Incorrect: LOW offgas flow is incorrect. Plausible:

Incorrect: Third stage is incorrect. Plausible:

Level: RO Exam

Lesson Plan Objective: AIRREMEOI 1



Reference(s): HC.OP-SO.CG-0001


KA: 239001 .K1.07 Knowledge of the physical connections andor cause effect relationships between MAIN AND REHEAT STEAM SYSTEM and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) K1.07 Offgas system 3.1 3.1



1 Pt(s) The plant is operating at 97% power limited by turbine back pressure. The “ B primary condensate pump trips. Which of the following is the correct automatic response of the plant?

A. RFP’s are limited to 4961 rpm; Recirculation System receives an intermediate runback signal

RFP’s are limited to 4961 rpm; Recirculation System is unaffected.

B.

C. RFP’s are limited to 4816 rpm; Recirculation System is unaffected.

D. RFP’s are limited to 4816 rpm; Recirculation System receives a full runback signal.

_____-____--------------------------------------~---------------------- Distracter Analysis:

A. Incorrect: 4816 RPM from DFCS and full runback on Recirc see

Plausible: Incorrect: 4816 RPM from DFCS and full runback on Recirc see

Plausible: C. Incorrect: Full runback on Recirc; see HC.OP-SO.BB-0002.

Plausible: D. Correct: IAW HC.OP-SO.BB-0002 and HC.OP-SO.AE-0001 Rev

48 Section 3.3.3 & 3.3.4.

HC.OP-SO.BB-0002 and HC.OP-SO.AE-0001.

B. HC.OP-SO.BB-0002.

Level: RO Exam

Lesson Plan Objective: FWCONTEOI 5 (R) From memory, describe the three possible RFP runback signals including conditions, setpoints and time delays if applicable, IAW the Feedwater Control System Lesson Plan.




Reference( s): HC.OP-S0.b.E-0001 Rev. 48 Section 3.3.3 & 3.3.4 (Revision with RPMchunge us provided by Archie Fuulkner) HC.OP-SO.BB-0002

KA: 259001 .K4.10 Knowledge of electrical power supplies to the following: (CFR: 41.7) Feedpump runbacks: Plant-Specific 3.1 3.4



1 Pt(s) The plant is in OPCON 1 at 20% power with the following offgas system indications:

Feed Gas Cooler Condenser Outlet Temperature is 250°F Preheater Outlet Temperature is 300'F Recombiner Outlet Temperature is 750°F

Which one of the following describes the status ofMain Condenser backpressure and the availability of the Mechanical Vacuum Pump (MVP)?

Under the current conditions, Main Condenser backpressure.. .

A. will rise. The MVP may NOT be placed in service.

B. will rise. The MVP may be placed in service if needed.

C.

D.

will remain stable. The MVP may be placed in service if needed.

will remain stable. The MVP may NOT be placed in service.

Distracter Analysis: A. Correct: The offgas train will isolate when Feed Gas Cooler

Condenser Outlet Temperature reaches 200°F resulting in a rise in Main Condenser backpressure. The MVF' may NOT be placed in service with reactor power above 5%. Incorrect: The MVP may NOT be placed in service with reactor power above 5%. Plausible: Incorrect: The offgas train will isolate when Feed Gas Cooler Condenser Outlet Temperature reaches 200'F resulting in a rise in Main Condenser backpressure. The MVP may NOT be placed in service with reactor power above 5%. Plausible: Incorrect: The offgas train will isolate when Feed Gas Cooler Condenser Outlet Temperature reaches 200°F resulting in a rise in Main Condenser backpressure. Plausible:

B.

C.

D.


Level: RO Exam

Lesson Plan Objective: GASRWOE013 (R) Concerning Recombiner TripdIsolations, determine the following IAW available control room references: a. The signals whch will cause a recombiner isolation b. Valves affected by an isolation signal

Source: Ne%


Reference( s): HC.OP-AB.BOP-0006 rev. 9

KA: 271000.K3.01 Knowledge of the effect that a loss or malfunction of the OFFGAS SYSTEM will have on following: (CFR: 41.5 / 45.3) K3.01 Condenser vacuum 3.5 3.5




The plant is operating at 100 percent power A Loss of Offsite power occurs coincident with a LOCA Drywell pressure is 5 psig and rising “A” Emergency Diesel Generator fails to start

SELECT the total FRVS recirculation flow 3 minutes after the LOCNLOP began. Assume NO operator action.

A. Ocfm

B. 90.000 cfm

C. 120.000 cfm

D. 180,000 cfm

_________________________l_l____________---------------------------


A.

B.

Incorrect: 4 fans start for a total of 120,000 cfm. Plausible: Would be true if there was no LOCA initiation signal. Incorrect: 4 fans start for a total of 120,000 cfm. Plausible: Would be true if 3 fans started. Need to h o w power supplies and system response to an initiation signal. Correct: “A” EDG powers “A” &“E” FRVS Recirculation fans. B, C, D, & F will start on the high drywell pressure signal. Each fan develops 30,000 cfm recirculation flow, so 4 X 30,000 cfm equals a total of 120,000 cfm. Incorrect: Only 4 fans start for a total of 120,000 cfm. Plausible: Would be true if all fans had power and NO operator action. Normally 6 fans start for 180,000 cfm and then operators secure 2 fans.

C.

D.

Level: RO Exam

Lesson Plan Objective: SECCONE012 (R) Given a set of conditions and a drawing of the controls, instrumentation and/or alarms located in the main control room, identify the status of the Secondary Containment by evaluation of the controls/ instrumentatiodalanns.


Source: Mod HC Bank Q68889


Reference(s): HC.OP-SO.GU-0001

KA: 288000.K4.02 Knowledge of PLANT VENTILATION SYSTEMS design feature(s) and/or interlocks which provide for the following: (CFR: 41.7) K4.02 Secondary containment isolation. 3.7 3.8



1 PYS) WHICH of the following describes the setpoint when the Reactor Building- to-Torus Vacuum Breakers b u t t d y isolation valve opens and the valve type?

When Reactor Building pressure exceeds the Suppression Chamber pressure by:

A. 0.25 psid, a pneumatic actuator opens the butterfly isolation valve.

B. 0.50 psid, a pneumatic actuator opens the butterfly isolation valve.

C.

D.


A.

0.50 psid, a motor operator opens the butterfly isolation valve.

0.25 psid, a motor operator opens the butterfly isolation valve. ______--_-______1-______1_______1____1__------------------------------------

Correct: T.S surveillance 4.6.4.2.b.2.a) states valve opening not to exceed 0.25 psid. M-57-1 shows the butterfly isolation valves have pneumatic actuators

Plausible: Incorrect: Valve opening setpoint is 0.25 psi& the valve has a pneumatic actuator. Plausible: Incorrect: The valve has a pneumatic actuator. Plausible:

B.

C.

Incorrect: The valve opening setpoint is 0.25 psid.

D.

Level: RO Exam

Lesson Plan Objective: PRICONE003 Summarize the basic construction and €unction of the following Primiuy Containment components: Vacuum Breakers, etc.



Refaence(s): M-57-1 & Lesson Plan


KA: 29000f.K5.01 Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT : (CFR: 41.5 / 45.3) K5.01 Vacuum breaker operation: BWR-4. 3.3* 3.4*



1 Pt(s) The reactor is at 90% power during a control rod exchange. The intermediate control rod configuration resulted in operations with MCPR equal to the limit specified in the COLR.

What MUST be done prior to withdrawing the next control rod?

A. Core flow must be reduced until the rod line is lowered.

B.

C.

D.


A RBM Channel Functional Test must be performed.

The reactor engineer must revise the RWM rod sequence.

The Pressure Regulator setpoint must be lowered. _---------_-_-------------------------------------------------------~---

A. Incorrect: Th~s may cause the Thermal Limit to exceed the COLR setpoint, since MCPR is flow dependent. Plausible: MCPR is flow dependent.

Specification 4.1.4.3 .b. Incorrect: This is an option to correct the pattern but it is not a requirement, but the TS specifically require a RBM functional prior to rod withdrawal while on a limiting pattern. Plausible: This is an option to correct the problem. Incorrect: Changing the regulator setpoint will change void concentration and a change in the void concentration may cause the thermal limit to lower or rise based on rod pattern. In addition, TS specifically requires a RBM functional prior to the next rod withdrawal while on a limiting pattern. Plausible: Could correct the problem depending on the rod pattern.

B. Correct: IAW HC.OP-10.22-0006 Caution 5.1.6 and Technical

C.

D.

Level: RO Exam

Lesson Plan Objective: IOP006E005 (R) Analyze plant conditions and parameters to determine if plant operation is in accordance with the POWER CHANGES DURING OPERATION Integrated Operating Procedure, supporting System Operating Procedures and Technical Specifications.




Reference(s): HC.OP-IO.ZZ-0006 rev 37 - step 5.1.6 Caution Prior to control rod withdrawal, if the reactor is operating on a LIMITING CONTROL ROD PATTERN, then Tech Spec 4.1.4.3.b shall be complied with.

Technical Specifications define a LIMITING CONTROL ROD PATTERN as a pattern that results in the core being on a thermal hydraulic limit; i.e., operating on a limiting value for APLHGR, LHGR or MCPR. Technical Specification 4.1.4.3.b requires that a Channel Functional Test for the Rod Block Monitor be performed prior to control rod withdrawal when the reactor is operating on a Limiting Control Rod Pattern.

KA: G2.1.12 2.1.12 Ability to apply technical specifications for a system. (CFR: 43.2/43.5/45.3)IMPORTANCERO2.9 SR04.0



1 Pt(s) Which one of the following is correct regarding power excursions of greater than 100% rated power with the Cross flow feedwater flow correction factor applied to the heat balance?

A. Permitted as long as NO thermal limit is exceeded and the twelve hour average power does NOT exceed 100%.

Permitted as Long as power does NOT exceed 102% for more than 15 minutes and the average power for any twelve hour shift does NOT exceed 100%.

Permitted as long as power does NOT exceed 102% and the average power for any twelve hour shift does NOT exceed 100%.

B.

C.

D. NOT permitted unless the SM invokes 10CFR50.54 X.


A. Incorrect: Power limitations above 100% are based on time, not thermal limits. 12 hour shift average and 12 hour average are not the same. Plausible: Incorrect: It is permissible to exceed this limit by up to 2% for as long as 15 minutes but in no case should 102% be exceeded Plausible: Correct: The average power level over any 12 hour shift shall not exceed licensed rated thermal power. It is permissible to exceed this limit by up to 2% for as long as 15 minutes but in no case should 102% be exceeded. Further guidance is as follows: 101 % for 30 minutes, 100.5% for 1 hr. etc. Incorrect: Improper use of 50.54.X. Plausible:

B.

C.

D.

Level: RO Exam

Lesson Plan Objective: IOP006E003 (R) Apply Precautions, Limitations and Notes while executing the POWER CHANGES DURING OPERATION Integrated Operating Procedure.




Reference(s): HC.0P-IO.ZZ-0006, rev 37 Step 3.1.6

KA: G2.1.10 2.1.10 Knowledge of conditions and limitations in the facility license. (CFR: 43.1 /45.13) 2.7 3.9



You and the Rx Building NE0 have been directed to reposition and independently verify EG-V.541, RHR HX BE205 SACS RTN ISLN VLV, located in the B RHR Heat Exchanger Room. The Heat Exchanger Room is posted as a “Locked High Rad Area.”

Which of the following is an acceptable means for Independent Verification?

1 Pt(s)

A. Observe change in SACS flow using CRIDS and Hardwire indications in the Main Control Room.

B. Verification using remote position Indication in the Main Control Room,

C. NO Independent Verification is required due to high cumulative radiation exposure.

After Initial positioning and locking of the valve, remove the lock count the turns to close, reposition valve and relock.

D.

_---___-___-----_-__--------------------------------------------------- Distracter Analysis:

A.

B.

Correct: IAW SH.OP-AP.ZZ-103, rev 12, attachment 12, section 1.7.b. Incorrect: Remote position indication is not available for this valve. Plausible: Incorrect: Some form of IV is always required. Plausible: Incorrect: high dose would dictate other means for verification, improper verification of a locked valve this would be a dual verification. Plausible:

C.

D.

Level: RO Exam

Lesson Plan Objective: ADMPROl03CE001 From Memory determine the requirements for Independent Verification LAW SH.OP-AP.ZZ-0103 and HU-AA-101.




Reference(s): SH.OP-AP.ZZ-103, rev 12, Attachment 12, Section 1.7.b

KA: G2.1.29 2.1.29 Knowledge of how to conduct and verify valve lineups. (CFR:41.10/45.1 /45.12)IMPORTANCERO3.4SRO3.3



Given the following conditions: 1 Pi@)

Reactor power is 30% "D" Diesel Generator is paralleled with the bus for surveillance testing "D" DG load is 2200KW A system malfunction causes an inadvertent initiation signal in the "D" Core Spray logic

Which statement below describes the response of the "D" DG to the conditions above?

A. The Normal bus supply breaker trips open and the diesel generator supplies the 10A404 bus.

The generator output breaker will trip open but the diesel remains running.

The diesel generator remains in parallel operation.

The generator output breaker will trip. The diesel shuts down and must be manually restarted.

B.

C.

D.

_-_1-____-___-__--___l_________________l---------------------------------


A. Incorrect: The EDG breaker will trip, and the normal supply breaker will carry the bus. Plausible: Correct: The following response occurs when a LOCA signal is received while the DG is operating in Test: the DG Output Breaker Trips, the DG remains running, the voltage regulator shifts to ISOC, and the Test Lockout Relay trips are disabled. If offsite power is available, the bus remains powered from the normal or alternate feed. If breaker closure permissives are met, the DG breaker will close to restore power. Incorrect: The EDG breaker will trip, the EDG will remain running, and the governor and regulator will shift to isochronous. Plausible: Incorrect: The EDG breaker will trip; however, the EDG will remain running. Plausible:

B.

C.

D.


Level: RO Exam

Lesson Plan Objective: 1EACOOEOlO (R) From memory, summarizehdentify the response of the diesel generators when running in the test mode and a LOPiLOCA signal is received.

Source: HC Bank Q5SOS4


Reference(s): Lesson Plan

KA: G2.2.12 2.2.12 Knowledgeofsurveillanceprocedures. (CFR: 41.10/ 45.13) IMPORTANCE RO 3.0 SRO 3.4



Given the following initial conditions: 1 Pt(s)

HPCI is currently operable. The Unit is in OPCON 1 at 100% power.

1AVH-209 HPCI Room Cooler control switch is in “AUTO LEAD”. 1BVH-209 HPCI Room Cooler control switch is in “AUTO”

Later in the shift,

A Maintenance Electrician reports to the Control Room and requests permission to place 1BVH-209 HPCI Room Cooler control switch in “STOP” for about 30 minutes to take voltage readings on the control switch contacts. The Electrician states that he will remain near the switch while performing his work and that he can restore his work and place the control switch back to “AUTO’ in less than 15 minutes in the event of a HPCI start. The CRS gives the Electrician permission to perform the work. The Electrician places the 1BVH-209 HPCI Room Cooler control switch in “STOP” at 1O:OO (the current time).

What is the status of HPCI?

A. HPCI is INOPERABLE since BOTH Room Coolers are NOT operable.

HPCI remains operable since Room Coolers are NOT required to support HPCI operability.

HPCI remains operable since BOTH Room Coolers are operable to support HPCI.

HPCI remains operable since ONE Room Cooler is operable to support HPCI.

B.

C.

D.



A.

B.

C.

D.

Incorrect: HPCI is operable since IAVH-209 is unaffected by the work and only one coolsr is required to be operable to support HPCI operability, Plausible: Would be hue if both coolers were required to support HPCI operability. Incorrect: At least one room cooler is required to support HPCI operability. Plausible: Some plants have performed analyses to prove that room coolers are not required for HPCI operability, so the answer is plausible. Incorrect: 1BVH-209 is NOT operable. Credit can NOT be taken for manual action to restore the cooler in this case since there was no mention of an operability determination to support manual action in the stem. In addition, it is not likely that the Electrician would qualify as a “dedicated” operator for compensatory action purposes. Plausible: Would be correct if the credit were taken for manual action. This distracter might be attractive to someone believing that both coolers are required for HPCI operability. Correct: One cooler is required to be operable to support HPCI in accordance with HC.OP-AP.ZZ-0108. IAVH-209 is unaffected by the work and remains operable; therefore, HPCI is operable.

Level: RO Exam

Lesson Plan Objective: HPCIOOE006 (R) From memory, summarize the interrelationship(s) between the HPCI System and any of the following IAW control room references: RBVS, etc.

Source: New


Reference(s): HC.OP-AP.ZZ-0108 rev. 28

KA: G2.2.24 2.2.24 Ability to analyze the affect of maintenance activities on LCO status. (CFR: 43.2 / 45.13) IMPORTANCE RO 2.6 SRO 3.8



1 Pt(s) Given the following conditions: 0 The Hope Creek Reactor is being shutdown to comply with Tech Specs

Rx power is 15% by APRMs RWM Group 34 has 4 of its 18 rods inserted to the target position of Notch 12 A rod is selected from RWM Group 16 0

The selected control rod can:

A. NOT be moved via RMCS; OUT OF SEQUENCE will be displayed on the RWM.

NOT be moved via RMCS; SELECT ERROR wiU be displayed on the R W .

B.

C . be moved via RMCS; SELECT ERROR will be displayed on the RWM.

D. be moved via RMCS; WITHDRAW ERROR will be displayed on the RWM.


A. Incorrect: OUT OF SEQUENCE is displayed under the following conditions: (1) >I IE, (2) 21 WE, (3) 1 IE and 1 WE, or (4) 1 IE or WE that is >2 notches. In this case there are no INSERT or WITHDRAW errors, only a SELECT ERROR. Plausible:

latched RWM Step (Pull Sheet Group) will result in a SELECT ERROR (SE), which results in an INSERT BLOCK (IB) and WITHDRAW (WB) block for that control rod. Incorrect: The rod cannot be moved due to INSERT WITHDRAW blocks being applied whenever there is a SELECT ERROR. Plausible: Incorrect: The rod cannot be moved due to INSERT gncJ WITHDRAW blocks being applied whenever there is a SELECT ERROR. Plausible:

B. Correct: Selection of a control rod that is in the currently

C.

D.


Level: RO Exam

Lesson Plan Objective: RODMINEOO3 (R) Given a labeled drawing of, or access to, the RWM Operator Display on 10C65 1, or the RWM Computer Display (in the Computer Room): a. Explain the function of each indicator. b. Assess plant conditions, which will cause the indicator to light or extinguish. c. Determine the effect of each control on the Rod Worth Minimizer. d. Assess plant conditions or permissives required for the control switches/pushbuttons to perform their intended functions. Source: HC Bank Q54169


Re ference(s): HC.OP-IO.ZZ-0003, Section 5.2, Approach to Criticality HC.OP-SOSF-0003, Section 5.3, Operations with InsertiWithdraw Errors

KA: G2.2.33 2.2.33 Knowledge of control rod programming. (CFR: 43.6) IMPORTANCE RO 2.5 SRO 2.9



1 Pt(s) Given the following conditions;

Reactor power is 100%. 0 Maintenance is being performed adjacent to the Main Steam Lines.

Which of the following describes the operation of the HWCI System regarding ALARA?

Per HCOP-10.ZZ-0006, Power Changes during operation; the Hydrogen and Oxygen Chemistry (HWCI) System should be

A. shutdown for eight hours prior to maintenance to reduce plant background radiation.

shutdown for as short a time as possible to prevent IGSCC.

left in service to prevent IGSCC.

left in service to reduce plant background radiation.

B.

C.

D.


A. Incorrect: The HWCI outages should be infrequent, and as short as possible, and preferably shorter than 8 hours. In addition, the dose rate drops of as soon as H2 water chemistry is secured so there is no need to wait 8 hours. Plausible: The procedure states that HWCI outages should be LIMITED to 8 hours, so it is plausible for someone to incorrectly recall the significance of the 8 hour time frame.

B. Correct: lAW HC.OP-IO.ZZ-0006 rev 37 section 3.2.6 states Background Radiation exposure within the turbine &reactor buildings will rise because of Hydrogen Water Chemistry (HWCI) system operation. If ALARA conditions warrant, HWCI should be shut down to perform maintenance or operations. The HWCI outages should be infrequent and as short as possible and preferably shorter than 8 hours. Intergrannular stress corrosion crack (IGSCC) growth may reinitiate after about 8 hours without HWCI. Incorrect: If ALARA conditions warrant, the HWCI System should be shutdown to perform maintenance or operations. Plausible. In order to prevent IGSCC, it is desirable to leave HWCl in service.

C.


D. Incorrect: Background radiation exposure within the turbine and reactor buildings will increase because of Hydrogen and Oxygen Chemistry (HWCI) System Operation Plausible: It is plausible that someone could have the misconception that the purpose of HWCI is to reduce background rad levels.

Level: RO Exam

Lesson Plan Objective: IOP006E002 (R) Explain the Precautions, Limitations and Notes listed in the POWER CHANGES DURING OPERATION Integrated Operating Procedure.



Reference(s): HC.OP-10.ZZ-0006 rev 37 section 3.2.6

KA: G2.3.10 2.3.10 Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure. (CFR: 43.4 / 45.10) IMPORTANCE RO 2.9 SRO 3.3



1 Pt(s) A new 25 year old employee with NO prior dose is limited to an initial whole body dose control level of current year, AND the dose control level may be increased to as high as

PRIOR to exceeding the initial whole body dose control level.

A. 2000 ; 2750

B. 2000 : 3000

m r d y r TEDE for the

m r d y r TEDE for the first extension, with the proper approvals,

C. 4000 : 4750

D. 4000 ; 5000


A. Incorrect: Initial level is correct, but the dose may be extended to as high as 3000 mrendyr TEDE. Plausible: Initial level is correct. An extension could be granted to 2750; however, the first extension can go as high as 3000. Correct: IAW NCNA-AP-ZZ-0024 rev. 13 Attachment 1. Incorrect: The initial is 2000 mredyr TEDE and the first extension is to as high as 3000 mredyr TEDE. Plausible: Correct values after the SECOND extension. Incorrect: The initial is 2000 mredyr TEDE and the first extension is to as high as 3000 m r d y r TEDE. Plausible: 4000 is correct control level after the SECOND extension and 5000 is the federal annual limit.

B. C.

D.

Level: RO Exam

Lesson Plan Objective: NOH04ADM024C-04 Given a set of exposure conditions Identify the personnel responsible for approval of the following dose extension: a. Yearly Dose Extension b. C. Lifetime Dose Extension

Declared Pregnant Women Dose Extension

IAW NC.NA-AP.ZZ-0024


Source: New

Level of knowledge: Mmory

Reference(s): NC.NA-AP.ZZ-0024 rev. 13 Attachment 1

KA: G2.3.4 2.3.4 Knowledge of radiation exposure limits and contamination control i including permissible levels in excess of those authorized. (CFR: 43.4 i 45.10) IMPORTANCE RO 2.5 SRO 3.1



A transient is in progress and you need to communicate with the crew regarding plant status, set priorities, and discuss what is coming up next.

What type of verbal communication is appropriate for this situation IAW OP-AA-104-101?

1 Pt(s)

A. Pre-job Brief

B.

C. Crew Update

D. Summary Brief


A.

B.

C.

Heightened Level of Awareness (HLA) Brief

____________-__-____--------------------------------~---------

Incorrect; Used during normal non transient conditions. Plausible: Incorrect: Used during normal non transient conditions. Plausible: Incorrect: Used for quick communications (1 0 to 15 second duration) for procedure transitions, announcements of major plant status changes, etc. Plausible: Correct: IAW OP-AA-104-101. Usedto briefthecrew during transient conditions (1 to 3 minute duration) for the types of communications given in the stem.

Level: RO Exam

D.


Source: New


Reference(s): OP-AA-104-101

KA: G2.4.15 2.4.15 Knowledge of communications procedures associated with EOP implementation. (CFR: 41.10 / 45.13) IMPORTANCE RO 3 .O SRO 3.5




The plant is operating at 15% power, s t m p in progress A SINGLE control rod SCRAMJust occurred as a result of a troubleshooting error whle investigating a SCRAM test switch problem.

Which of the following groups of alarms are consistent with this condition?

A. CRD ACCUM TROUBLE (C6-D4) ROD DRIFT (C6-E3) SCRAM AIR HEADER LOW PRESSURE (CRIDS A3016)

B. ROD DRIFT (C6-E3) ROD OUT MOTION BLOCK (C6-D3) SCRAM AIR HEADER LOW PRESSURE (CRIDS A3016)

C. CRD ACCUM TROUBLE (C6-D4) ROD DRIFT (C6-E3) ROD OUT MOTION BLOCK (C6-D3)

D. CRD ACCUM TROUBLE (C6-D4) ROD OVERTRAVEL (C6-F3) ROD OUT MOTION BLOCK (C6-D3)

____________________________________L_I_----------------------------------------


A. Incorrect: SCRAM AIR HEADER LOW PRESSURE will not come in on a single rod scram. Plausible: 2 out of 3 are correct. SCRAM AIR HEADER LOW PRESSURE comes in on a full scram.

B. Incorrect: SCRAM AIR HEADER LOW PRESSURE will not come in on a single rod scram. Plausible: 2 out of 3 are correct. SCRAM AIR HEADER LOW PRESSURE comes in on a full scram. Correct: CRD ACCUM TROUBLE will come in due to low accumulator pressure following the rod scram. ROD DRIFT will come in since the rod is moving without an RMCS command. ROD OUT MOTION BLOCK comes from the RWM andor SDV not drained. Incorrect: ROD OVERTRAVEL is indicative of an uncoupled control rod NOT a scrammed rod.

C.

D.


Plausible: 2 out of 3 are correct. Rods travel slightly beyond 00 on a scram and settle at 00 once the scram is reset, so ROD OVERTRAVEL is a plausible misconception.

Level: RO Exam

Lesson Plan Objective: ABICOI EO01 Recognize abnormal indicationsialarms andor procedural requirements for implementing Control Rod.

Source: Mod HC Bank 455937

Level ofknowledge: Analysis

Reference(s): HC.OP-AB-IC-0001 rev. 4 - Alarms

KA: G.2.4.46 2.4.46 Ability to verify that the alarms are consistent with the plant conditions. (CFR: 43.5 / 45.3 / 45.12) IMPORTANCE RO 3.5 SRO 3.6




The Unit is in OPCON 1 at 100% power. Multiple CRIDS high temperature alarms are in for both the “A” and “B” Reactor Recirculation Pumps, and the affected temperature indications are trending up on the CRIDS display. The following points are trending up on both pumps:

o Pump Motor Oil high temperature o Upper & Lower Motor bearing high temperature o # 1 & #2 Seal Cavity high temperature

The plant is currently stable.

Which one of the following malfunctions could explain ths combination of indications, and under theses circumstances, what is the MAXIMUM amount of time that plant procedures permit continued Recirculation Pump operations?

A. Inadvertent isolation of Chilled Water flow to the Recirculation Pumps. Restore cooling within 10 minutes, or trip both Recirculation Pumps.

Inadvertent isolation of RACS cooling water flow to the Recirculation Pumps. Restore cooling within 10 minutes, or trip both Recirculation Pumps.

B.

C. Inadvertent isolation of Chilled Water flow to the Recirculation Pumps. Restore cooling within 20 minutes, or trip both Recirculation Pumps.

Inadvertent isolation of RACS cooling water flow to the Recirculation Pumps. Restore cooling within 20 minutes, or trip both Recirculation Pumps.

D.


A. Incorrect: Chilled Water normally supplies cooling to the Recirc Pump Motor AIR Coolers. A loss of Chilled Water would result in Motor WINDING hgh temperature indications and alarms NOT the indications presented in ths scenario. Plausible: Chllled water cools the motor air coolers, and it is often confused with RACS. RACS can provide backup cooling to drywell


chilled water loads and this contributes to the misconception. In addition, the maximum time specified is correct for a loss of RACS. Correct: RACS cooling supplies cooling water to the specified loads. HC.OP-AB.COOL-003, “Reactor Auxiliary Cooling” requires that the Recirc Pumps be tripped if RACS cooling CANNOT be restored within 10 minutes. Incorrect: Chilled Water normally supplies cooling to the Recirc Pump Motor AIR Coolers. A loss of Chilled Water would result in Motor WINDING high temperature indications and alarms. In addition, the maximum time specified is incorrect. Plausible: Chilled water cools the motor air coolers, and it is often confused with RACS. RACS can provide backup cooling to drywell chilled water loads and this contributes to the misconception. Incorrect: The maximum time specified is incorrect. Plausible: RACS is correct.

B.

C.

D.

Level: SRO Exam CFR 55.43(5)

Lesson Plan Objective: RACSOOE019 Explain the effect that a loss of RACS has on the Reactor Recirculation System, including any time restraints that may or may not be placed on operation of the Reactor Recirculation pumps. IAW available control room references

Source: New


Reference(s): Lesson Plan NOHOIRECIRC, “Recirculation System” Lesson Plan NOHOl RACSOOC, “Reactor Auxiliary Cooling” Lesson Plan NOHOl TBCWOO, “Turbine Building Chilled Water” HC.OP-AB.COOL-003, “Reactor Auxiliary Cooling”

KA: 295018.A2.03 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER (CFR: 41.10/43.5 /45.13)AA2.03 Cause forpartid or complete loss 3.2 3.5



Technical Specification LCO 3.6.1.7 states that drywell average air temperature shall not exceed 135°F.

The basis for this LCO is.. ...

1 Pt(s)

A. to ensure that containment peak air temperature does not exceed the design temperature during LOCA conditions.

to ensure that containment peak air temperature does not exceed the Tech Spec DryweU Temperature Safety Limit during LOCA conditions.

B.

C. to ensure that the containment can absorb the associated decay and structural sensible heat released during a reactor coolant system LOCA blowdown.

to ensure that RPV level instrument reference legs do not flash during a reactor coolant system LOCA blowdown.

D.


A. B.

Correct: Reference Tech Spec Basis for LCO 3.6.1.7. Incorrect: There is no Tech Spec Safety Limit for drywell temperature. Plausible: There are Tech Spec Safety Limits for the fuel barrier and RPV barrier, so it would be a plausible misconception for someone to think that there would be a Tech Spec Safety Limit for Drywell pressure and or temperature. In addition, there is a Limiting Safety System Setting (LSSS) for drywell pressure. Incorrect: This is an excerpt from the Depressurization Systems LCO Basis 3.6.2 and it is related to the suppression pool water heat capacity. It may be a partially true statement, but it is not the basis for the dIywell air temperature. Plausible: This is true regarding suppression pool water temperature which is part of containment. Incorrect: Flashing is not an issue at 135"F, and flashmg is not discussed in the Tech Spec Bases. Plausible: Drywell temperatures above saturation can cause flashing, and it is a concern in the EOPs.

C .

D.

Level: SROExam CFR 55.43(2)


Lesson Plan Objective: TECSPCE009 Explain the bases for Hope Creek Generating Station Technical Specification Safety Limits and Limiting Safety System Settings. (SROISTA Only)

Source: New


Reference(s): Technical Specification Bases

KA: 295028.A2.03 2.2.25 Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. (CFR: 43.2)



1 Pt(s) At 18:00, you receive the following information at shift turnover: The Unit is in OPCON 4 and was shutdown earlier in the day to repair a known coil leak on Drywell Cooling Unit 1AVH212. “A” Loop RHR is in Shutdown Cooling with SDC flow being maintained at 7000 GPM. Reactor coolant temperature is 175°F at the “A” RHR HX inlet. RPV level is steady at 80 inches on the shutdown range. RPV pressure is steady at 15 psig on the feedwater CRT digital display. Reactor head vents are open. Both Reactor Recirculation Pumps are out of service.

Later in the shift: At 23:00, the Reactor Operator observed RPV pressure change from 15 psig to 18 psig over the last hour on the feedwater CRT digital display along with a noticeable rise in drywell cooler drain flow. Reactor coolant temperature is still steady at 175°F at the “A” RHR HX inlet.

Which one of the following explains the RPV pressure indication and Shutdown Cooling status?

A.

B.

C.

D.

The RPV is not steaming and the pressure indication is normal at the low end of instrument range. SDC flow is too low; however, core circulation is satisfactory since reactor coolant temperature is steady and well below 20O0F as measured at the RHR HX inlet.

The RPV is not steaming and the pressure indication is normal at the low end of instrument range. SDC indications are normal since reactor coolant temperature is steady and well below 200°F as measured at the RHR HX inlet.

The RPV is steaming. SDC flow is too low to ensure adequate core circulation and flow must be raised.

The RPV is steaming. RPV level is too low to ensure adequate core circulation and level must be raised.

How Creek 2007 JLT Exam

Distracter Analysis: A. Incorrect: Core circulation is NOT satisfactory with SDC flow less

than 10,000 GPM, and RHR HX inlet temperature may not be representative of the bulk coolant temperature, so temperature stratification may occur. RPV pressure instruments may not read exactly 0 psig; however, a 3 trend up over an hour may indicate that steaming is occurring. Plausible: SDC flow is Low and RHR HX inlet temperatures suggest adequate core circulation and coolant temperature less than 200°F. Incorrect: Core circulation is NOT satisfactory with SDC flow less than 10,000 GPM, and RHR HX inlet temperature may not be representative of the bulk coolant temperature, so temperature stratification may occur. RPV pressure instruments may not read exactly 0 psig; however, a 3 trend up over an hour may indicate that steaming is occurring. Plausible: RHR HX inlet temperatures suggest adequate core circulation and coolant temperature less that 200"F, so someone could infer that SDC indications are normal. Correct: SDC flow is too low to ensure adequate core circulation. The slight rise in RPV pressure along with increasing drywell drain flow indicates that there is temperature stratification in the RPV and that RCS coolant temperature may be above 212°F near the surface resulting in steaming through the RPV vents to the drywell. The RHR HX inlet temperature may not be indicative of bulk coolant temperature at reduced shutdown cooling flow (less than 10.000 GPM). Incorrect: Reactor level is maintained at 80 inches to ensure that a natural circulation path exists if forced circulation is lost, so it is at the correct value. Plausible: The slight rise in RF'V pressure is an indication that steaming may be occumng in the RPV. In addition, level is a parameter important to natural circulation.

Level: SRO Exam CFR 55.43(2) and CFR 55.43(5)

B.

C.

D.


Lesson Plan Objective: RHRSYSE009 Given plant problemslindustry events associated with the Residual Heat Removal System: a. Discuss the root cause of the plant problemiindustry event IAW

the associated plant problemslindustry event document. b. Discuss the HCGS design and/or procedural guidelines that

mitigateireduce the likelihood of the problemiindustry event at HCGS IAW the associated plant problemdindusky event document.

c. Discuss the “lessons learned” from this probldevent IAW the associated plant problemshndustry event document.

Source: New


Reference( s): HC.OP-SO.BC-0002, “Decay Heat Removal Operations” Lesson Plan NOHOIRHRSYS, “Residual Heat Removal”

KA: 295021.A2.06 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING : (CFR: 41.10 / 43.5 145.13) AA2.06 Reactor pressure 3.2 3.3




0

0

0

0

The Unit is in OPCON 1 at 100% power. A loss of all drywell cooling occurs. Drywell pressure rises to 1.4 psig and operators manually scram the reactor. RPV level dropped to a low of -10 inches and was restored to +30 inches with feedwater. Drywell pressure continues to rise to 2.0 p ig . All automatic actions occur as designed. HPCI injected and was immediately secured since level control was established with feedwater. Drywell cooling was restored several minutes later, all isolations were reset, and the plant was stabilized in OPCON 3.

Which one of the following is the MOST LIMITING reportahility requirement that applies the event described above?

A. 15 minute EMERGENCY notification to the states & counties.

B. 1 hour NON-emergency notification to the NRC Operations Center.

C. 4 hour NON-emergency notification to the NRC Operations Center.

D. 8 hour NON-emergency notification to the NRC Operations Center.


A. Incorrect: This event does not meet an EAL threshold; therefore, no EMERGENCY notifications are required. The high drywell pressure condition resulted from a loss of drywell cooling and was NOT indicative of a loss or potential loss of the RCS barrier. See the basis for EAL # 3.2.2.b. Plausible: There is an EAL for high drywell pressure due to a loss of the RCS barrier. If an actual RCS leak was the cause of the condition, then ALERT declaration would he required and that would require 15 minute emergency notification to the states and counties.


B. Incorrect: Does not meet the threshold for 1 hour notification. Plausible: 1 hour non-emergency notifications do exist, and in the past, ECCS injection to the vessel did require 1 hour notification. Correct: 4 hour NON-emergency notification to the NRC Operations Center due to the valid HPCI actuation and injection to the vessel and due to the manual RPS actuation in anticipation of the automatic scram on 1.68 psig. The HPCI initiation was valid since the signal was generated by the sensor which measured a physical system parameter that was at its set point. See RAL #11.3.1 and 11.3.2. Incorrect: The 4 hour report is the most limiting reportability requirement in this case. Plausible: The primary containment isolation on 1.68 psig qualifies for an 8 hour notification.

C.

D.

Level: SRO Exam CFR 45.3

Lesson Plan Objective: NOH04ADA120C-03 From memory, describe the actions for Operations Shift Management Review of an issue to determine the following:

If operability is required Determine Operability Reportability requirements

1Aw LS-"4-120.

Source: New


Reference(s): Hope Creek Event Classification Guide & Bases

KA: 295024.G2.4.30 2.4.30 Knowledge of which events related to system operationsistatus should be reported to outside agencies. (CFR: 43.5 /45.11)




The plant is operating at 100% power. All systems are aligned for normal operations. A fire started under Control Room console 10C65 1 causing a reactor scram. The Control Room has been evacuated because of extreme smoke conditions. The reactor has been depressurized to less than 80 psig with SRV’s and RCIC. “B” RHR is in Suppression Pool Cooling from the RSP.

Which one of the following describes the preferred Remote Shutdown system to be used to achieve Cold Shutdown, and what is the maximum cooldown rate that is permitted in accordance with HC.OP-IO.ZZ-0008, “Shutdown from Outside the Control Room”?

A.

B.

C.

D.

‘‘A’’ RHR in Shutdown Cooling at less than 100 ‘F/hr.

“A” RHR in Shutdown Cooling at less than 90 OF/hr.

“B” RHR in Shutdown Cooling at less than 100 OF/hr.

“B” RHR in Shutdown Cooling at less than 90 ‘Fhr.


A. Incorrect: “B” SDC is the preferred SDC loop to be used if at all possible as directed by the procedure. The procedure directs less than 90 “ F h . Plausible: Provisions are provided in the procedure to us “A” SDC, but the procedural actions are more complex and time consuming than the “B’ loop. The Tech Spec limit is less than 100 “Fihr. Incorrect: “B’ SDC is the preferred SDC loop to be used if at all possible as directed by the procedure. Plausible: Provisions are provided in the procedure to us “A” SDC, but the procedural actions are more complex and time consuming than the “B’ loop. The rate is correct. Incorrect: The procedure directs less than 90 “ F h . Plausible: “B” SDC is correct, and the Tech Spec limit is less than 100 “ F h .

B.

C.


D. Correct: HC.OP-IO.ZZ-0008, “Shutdown from Outside the Control Room” step 5.9.6 specifies that “B” RHR be secured from Suppression Pool Cooling and be placed in Shutdown Cooling mode. The procedure directs that the cooldown be maintained at less than 90 O F h .

Level: SRO Exam CFR 55.43(b)(5)

Lesson Plan Objective: REMSIDE002 Given a system that is controlled from, or is required to support the operation of the Remote Shutdown System, explain the function of the supporting system. IAW available control room references.

Source: Mod Hope Creek Bank (456470) Audit Exam 9/99


Reference(s): NOHOIREMSD, “Remote Shutdown” HC.OP-IO.ZZ-0008, “Shutdown from Outside the Control Room”

KA: 295016.A2.06 Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT : (CFR: 41.10 / 43.5 / 45.13) AA2.06 Cooldown rate 3.3 3.5



1 Pt(s) During a plant startup at 5% reactor power, the on-shift Chemistry Technician reports the following results with regards to the Standby Liquid Control Storage Tank:

NET TANK VOLUME - 4700 gallons % WEIGHT SOLUTION CONCENTRATION - 13.7% SLC Tank temperature is 80°F

Which of the following describes the SLC system status and the basis for that conclusion?

A. The SLC system is operable since there is sufficient boron to bring the reactor from 5% power to a cold, Xenon free shutdown assuming that withdrawn control rods remain fued in the 5% power pattern.

The SLC system is NOT operable since there is insufficient boron to bring the reactor from full power to a cold, Xenon free shutdown assuming that withdrawn control rods remain fued in the rated power pattern.

The SLC system is NOT operable since the solution is OVER saturated, so boron may precipitate out of solution such that there will not be enough boron to overcome the reactivity insertion rate due to cooldown and Xenon decay.

The SLC system is operable since the solution is UNDER saturated, so boron will NOT precipitate out of solution so there will be enough boron to overcome the reactivity insertion rate due to cooldown and Xenon decay.

B.

C.

D.


A. Incorrect The Reactor is in OPCON I, so SLC must meet the Tech Spec operability requirements regardless of the low power level. Plausible: In reality, there is probably enough SLC to shutdown the reactor from 5% initial power. Correct: The SLC volume is too low for the existing concentration, so there is insufficient boron to ensure that the reactor can be brought from full power to a cold, Xenon free shutdown assuming that withdrawn control rods remain fixed in the rated power pattern.

B.


C. Incorrect: The system is inoperable based on the total amount of boron currently available to inject. Precipitation should not be a problem with tank temperature at 80°F. Plausible: The system is inoperable. Precipitation can be a problem at less than 70'F. In addition, some of the words regarding Xenon and reactivity insertion rate were pulled from the Tech Spec Bases. Incorrect: The S u 3 system is NOT operable. Plausible: An under saturated solution would be less susceptible to precipitation, so if someone thought precipitation was the issue, and if they concluded the solution was under saturated, then they could conclude the system was operable. In addition, some of the words regarding Xenon and reactivity insertion rate were pulled from the Tech Spec Bases.

Level: SRO Exam CFR 55.43@)(2)

Lesson Plan Objective: SLCSYSE018 Provided access to Technical Specifications, analyze conditions to determine if SLC System parameters are within the Acceptable Operating Region of the Sodium Pentaborate Solution Volumei Concentration Requirements Graph, IAW HCGS Technical Specifications.

Source: Mod HC Bank Question Q56443


Reference(s): Tech Specs &bases

D.

KA: 295037.G2.1.33 2.1.33 Ability to recognize indications for system operating parameters which are enby-level conditions for technical specifications. (CFR: 43.2 i 43.3 / 45.3) SRO 4.0



1 Pt(sj Given the following:

The Unit was operating at 80% power. A Main Steam line rupture occurred in the turbine building. A Group I Primary Containment Isolation signal was received on High Main Steam Line Flow. The MSN’s failed to completely isolate the leak.

Reactor level dropped to -45 inches, and level was restored to +30 inches with HPCI & RCIC. Primary & Secondary containment parameters remained normal. Personnel reported a large steam plume in the turbine building upstream of the turbine stop valves The reactor has been depressurizing through the steam line rupture for the last 75 minutes.

The reactor automatically scrammed.

Current conditions (75 minutes after the event began): Reactor pressure is 200 psig and dropping slowly. The steam line leak is still in progress. Reactor level is +30 inches with condensate injecting for level control. HPCI & RCIC have been secured to slow the depressurization rate, and both systems were realigned for automatic operations. Primary & Secondary containment parameters are normal. Health Physics has just provided a Dose Assessment projection based on Plant effluent sample analysis as follows:

o 3.4E-01 mRem TEDE 4-day dose at the MEA o 1.6E-01 &em Thyroid-CDE

Which one of the following describes the highest event classification and the Emergency Operating Procedure(s) that were or are required to be entered to mitigate the event?

A. General Emergency. EO.ZZ-0101 & EO.ZZ0104.

B. Site Area Emergency. EO.ZZO101 only.

C. Alert. EO.ZZO101 only.

D. Unusual Event. EO.ZZOlOl& EO.ZZ-0104.



A.

B.

C .

D.

Incorrect: The threshold was not reached for GE classification, and EOP-104 entry was not required. See justification for B. Correct: SAE declaration required due to a Main Steam Line break outside containment with a failure of the MSIV's to isolate leak with a downstream pathway to the environment. EOP-0101 was required to be entered on low reactor level; however, EOP-0104 entry wadis NOT required since the dose projection has only reached the Unusual Event level based on TEDE. EOP-0104 entry is required when there is a gaseous radioactive release at the ALERT threshold. The SAE declaration was based on the loss of the RCS and containment barriers NOT the radioactive release, so EOP-104 entry wadis not required. Incorrect: SAE threshold was reached. See justification for answer B. Incorrect: SAE threshold was reached & EOP-104 entry was not required. See justification for answer B. Plausible: UE is correct just for the rad release, but ths was not the highest EAL for the event.


Lesson Plan Objective: ABCNT4E001 Recognize abnormal indications/alms and/or procedural requirements for implementing Radioactive Gaseous Release.

Source: New


Reference(s): Hope Creek Event Classification Guide HC.OP-EO.ZZ-0101 & 0104.

KA: 295038.A2.03 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: (CFR: 41.10 / 43.5 / 45.13) AA2.03 +Radiation levels: Plant-Specific 3.1 3.9



While performing a panel walk down at 100% power you observe the following:

1 Pt(s)

“A” SPTMS Avg Suppression Pool temperature is 82°F. “B’ SPTMS Avg Suppression Pool temperature is 85°F. After checking the daily control console logs, you identify that the “B SPTMS Avg has risen 3°F over the last 24 hours. You then check the individual SPTMS channel inputs and identify that SPTMS channel B6 is reading 100°F while all of the other individual SPTMS channel inputs show indications that range between 80°F and 85°F. SPTMS channel B6 is located near Torus Azimuth 0” which is in close proximity to SRV tailpipe F013H. SRV F013H tailpipe temperature is currently reading 180°F and it has risen 40°F over the last 24 hours as recorded in the daily console

All other plant indications are normal. logs.

m c h one of the following is the correct action to take in response to these indications?

A. Enter HC.OP-AB.RF’V-0006, “Safety / Relief Valve” AND execute the immediate operator actions.

Declare “B” SPTMS inoperable AND enter the LCO for having less than the required number of operable Suppression Chamber Water Temperature channels.

Enter HC.OP-EO.ZZ-0102, “Primary Containment Control” due to high local Suppression Pool Water Temperature AND place Suppression Pool Cooling in service to lower the pool temperature.

Notify the SRV System Manager AND continue to monitor Suppression Pool Temperature and F013H SRV tailpipe temperature.

B.

C.

D.



A. Incorrect: This procedure is for an open SRV. There is indication of SRV lekage, but there are no other indications to conclude that the SRV is open (no acoustics alarm, no change in steam flow/feed flow, no change in pressure, etc.). Immediate actions would require reducing power and cycling the SRV control switch which would cause the SRV to open and possibly not reseat. Plausible: The temperature indications would be used to confirm that an SRV was open if there were other parameters or alarms indicating that an SRV opened. Incorrect: The SPTMS average temperature and the individual channel input are reflecting actual plant conditions, so they are NOT inoperable. Plausible: Tech Specs require 2 “channels” of SPTMS indication for average bulk temperature. In addition, someone could conclude that having one individual channel input reading high relative to the other inputs would render that channel inoperable. Incorrect: The EOP entry is based on average Suppression Pool temperature greater than 95°F. In this case, average temperature on the “B’ SPTMS is reading 85”F, so there is no emergency or reason to enter the EOP. The operators may elect to start Suppression Pool Cooling to lower the local temperature, but that can be done using normal plant procedures rather than Emergency Operating Procedures. Plausible: Local temperature is actually high, and starting suppression pool cooling would lower the temperature. Correct: The indications suggest that the SRV has pilot or main seat leakage. The daily logs have a note that the System Manager should be contacted if there is a greater than 30’F rise over the normal or baseline temperature. A reference is not needed to determine the correct answer. The symptoms alone indicate that the local SP temperature is rising due to SRV leakage near the SP temperature detector. The operator would need several other diverse indications to conclude that the SRV was open, and none of those indications exist at this time. Notifymg the System Manager & increasing monitoring is the prudent and correct action to take based on the indications presented.

B.

C.

D.

Level: SRO Exam CFR 55.43@)(2) & (5)


Lesson Plan Objective: EOPl02E004: From memory. recall the reason why average suppression pool temperature is used for determining the entry condition and subsequent actions IAW the Primary Containment Control - Suppression Pool Lesson Plan. ABRPV6E001: Recognize abnormal indicationsialarms and/or procedural requirements for implementing SafetyRelief Valve.

Source: New


Reference(s): HC.OP-AB.RPV-0006, “Safety / Relief Valve” HC.OP-EO.ZZ-0 102, “Primary Containment Control” HC.OP-DL.ZZ-0003, “Log 3 Control Console Log Condition 1,2, & 3”

KA: 295013.A2.02 Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: (CFR: 41.10 / 43.5 / 45.13) AA2.02 Localized heatingistratification.3.2 3.5




A LOCA has resulted from a seismic event Reactor water level is -20 inches and rising Reactor pressure is 850 psig and slowly lowering Drywell Pressure is 3 1 psig and slowly rising Drywell temp is 275 "F and slowly rising Suppression Chamber pressure is 30 psig and slowly rising Suppression Pool water level is 77 inches "B" RHR Loop is in Drywell Spray The Main Condenser is NOT available All control rods are full in

Based on the above conditions, when is depressurization of the reactor required and with what method?

A.

B.

C.

Immediately using 5 ADS Valves

Immediately using Turbine Bypass Valves

When Drywell Press reaches 35 psig using Turbine Bypass Valves

When Drywell Temp reaches 310F using 5 ADS Valves D.


A.

B.

__________-__I______________l_l_________-----------------------------------

Correct: Immediately using 5 ADS Valves. Emergency depressurization must occur now for exceeding the PSP curve. Incorrect: Immediately using Turbine Bypass Valves. Emergency de-pressurization can no longer be anticipated. Emergency depressurization must occw now for exceeding the PSP curve. Incorrect: When Drywell Temp reaches 310DF using 5 ADS Valves. Emergency de-pressurization must occur now for exceeding the PSP curve. Emergency de-pressurization for high Drywell temperature does not occur until 3400F. Incorrect: When Drywell Press reaches 35 psig using Turbine Bypass Valves. Emergency de-pressurization can no longer be anticipated. Emergency de-pressurization must occur now for exceeding the PSP curve.

C.

D.



Lesson Plan Objective: E0102PE007 Given any step of the procedure, determine the reason for performance of that step andor predict expected system response to control manipulations prescribed by that step IAW the Primary Containment Control - Drywell Lesson Plan.

Source: HC Bank (456157) Audit Exam 1999


Refaence(s): HC .OP-E0.ZZ-01 02(Q) curve SCP-L

KA: 2950101.A2.02 2.4.6 Knowledge symptom based EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)




A large break LOCA has occurred inside the Drywell. Multiple equipment failures occurred. Drywell pressure is 15 psig. Steam cooling was required until water level was restored above TAF with Fire Water. The Containment H2102 Analyzers were placed in-service 2 hours ago. The containment H2102 Analyzers have just alarmed on High Drywell H2 Concentration and the trend is upward at 0.5 percent per hour. H2 concentration is at 2.1% and 0 2 concentration is 1.3%. The H2 Recombiners are currently NOT in service.

Which one of the following actions is required IAW HC.OP-EO.ZZ.0102?

A. Exit EOP-102 and enter SAG since core damage is in excess of what the Emergency Operating Procedures were typically designed to handle.

Place the H2 Recombiners in service since there are sufficient quantities of H2 & 0 2 to support effective recombination.

Exit EOP-102 and enter SAG since the lower H2 detonation limit has been reached and containment failure may occur with known fuel damage.

Vent containment via the Suppression Chamber to reduce Drywell H2 concentration since the Recombiners could initiate an H2 detonation at these concentrations.

B.

C.

D.

_____I___--__-______________I___________---------


A. Correct: EOP-IO2 PC/H-1 directs that the EOP be exited and SAG entered is H2 concentration exceeds 2%. In addition, the bases states that a 2% H2 concentration confirms fuel damage above 10 CFR 50.46 ECCS design requirements and beyond what the EOPs were typically designed to handle. Incorrect: With concentrations above 2%, the direction is to exit the EOP and transition to SAG. The Recombiners could have potentially been placed in service earlier in accordance with EOP-

B.


102, but the stem indicates that the recombiners are currently out of service. Now that H2 concentration is above 2%, EOP-102 must be exited. Recombiner operations will be governed by SAG at this point NOT EOP-102. Plausible: EOP-102 does direct that the Recombiners be placed in service if concentration is above 0.5% since H2 is present in sufficient quantities for the Recombiners to be effective. Incorrect: Correct action but wrong reason. H2 and 0 2 concentrations are well below the LEL for H2. Plausible: Action is correct. In addition, containment failure would be possible if there were a detonation. Incorrect: Venting is not an option in EOP-102 under the current conditions. Plausible: Sag may direct venting to control H2 concentration at some point. In this event, then Suppression Chamber venting would be a good idea to allow scrubbing in the pool and the humid environment and sprays could be used as a means to prevent or mitigate am H2 burn.


Lesson Plan Objective: E01 02PE004 Recall the reasons why the following are used for determining the entry condition and / or subsequent actions IAW the Primary Containment Control - Drywell Lesson Plan. a. Drywell Pressure b. Average Drywell Temperature c. H2 and 0 2 concentrations in the drywell

Source: Mod HC Bank Question (468880) NRC exam 2002

C.

D.


Reference(s): HC.OP-EO.ZZ.0102 & Bases

KA: 500000.A2.03 Ability to determine and/or interpret the following as they apply to HIGH CONTAINMENT HYDROGEN CONCENTRATIONS: (CFR: 41.10 / 43.5 / 45.13) EA2.03 Combustible limits for drywell 3.3 3.8



1 Ptb? Given the following:

0

0

0

The reactor is at 27% power. Power ascension is in progress following a major refueling outage. Main Turbine Generator bearing number 10 and 1 1 were replaced during the outage. TURBINE GENERATOR VIB HI alarm D3-C5 just annunciated. Number 10 bearing is at 10 mils and rising slowly. Number 11 bearing is at 9 mils and rising slowly.

Which one of the following describes the required immediate actions?

A. REDUCE Recirculation Pump speed to MINIMUM, and then LOCK the Mode Switch in Shutdown, and then TRIP the Main Turbine.

B. IMMEDlATELY TRIP the Main Turbine, and then LOCK the Mode Switch in Shutdown.

C. LOCK the Mode Switch in Shutdown, and then TRIP the Main Turbine.

REDUCE reactor power to less than 25%, and then TRIP the MAIN Turbine.

D.

______I___________________1_1_1_________--------------------------------------


A. Incorrect: This is NOT the immediate action for when vibration is greater than or equal to 8 mils on Number 11 or 12 bearing. Plausible: Ths is the retainment override for the condition when vibration is greater than or equal to 7 mils on Number 11 or Number 12 bearing; however, vibration is now at 8 mils. Incorrect: Wrong sequence. Would result in reactor high pressure and high neutron flux on the turbine trip. Plausible: Could confuse the sequence since there is a problem on the turbine requiring immediate action. Correct: With Number 11 bearing above 8 mils with power greater than 25%, the immediate action is to lock the mode switch in shutdown, then trip the turbine. Incorrect: Procedure does NOT permit a controlled power reduction with the given vibration levels.

B.

C.

D.


Plausible: It would be nice to do this since the turbine could be tripped while leaving the reactor online on bypass valves.

Level: SRO Exam

Lesson Plan Objective: ABBOP2E003 From memory, recall the Immediate Operator Actions for Main Turbine.

Source: New


ReferencHs): HCOP-AB.BOP-0002, “Main Turbine”

KA: 245000.A2.09 Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS; and @) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) A2.09 Turbine Vibration 2.5 2.8




A reactor startup from COLD SHUTDOWN is in progress. The point of adding heat has just been reached. IRM B is inoperable and was bypassed prior to startup. IRM H is observed by the operators to fail downscale, producing a DOWNSCALE alarm hut no INOP trip.

The reactor startup:

A. is NOT permitted by Technical Specifications. Only rod motion by scram is permitted.

B. may continue if IRM H is placed on Range 1. The IRM trip system with fewer operable channels than required must be placed in the tripped condition within 12 hours. The provisions of Technical Specification 3.0.4 are not applicable.

may continue but the mode switch cannot be placed in the RUN position until one of the inoperable IRMs is returned to service.

is NOT permitted by Technical Specifications. The control rods may be inserted in the reverse sequence to shutdown the reactor.

C.

D.

___l_____l___________l_l_____________l__ ______--___-__I-____----------------


A.

B.

Incorrect: Continued operation is permitted. See B. Also, normal rod motion is permitted. Correct: Continued startup is permitted. Regarding Rod Block Instrumentation: Placing on Range 1 bypasses the downscale rod block to allow or ensure continued rod withdrawal. Only required to have 6 operable channels per trip function, and 6 remain. Regarding RPS instrumentation: Tech Spec 3.3.1.a and table 3.3.1-1. Table3.3.1-1 statesforOpCon2need3 IRMspertrip system. 3.3.1.a states with the number of operable channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel (s) and/or that trip system in the tripped condition* within twelve hours. The provisions of Specification 3.0.4 are not applicable. 3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made when the conditions for the Limiting Condition for Operation are not met and the associated


ACTION requires a shutdown if they are not met within a specified time interval. Incorrect: Can place the mode switch in RUN since the RM’s are automatically bypassed with the mode switch in RUN. Incorrect: Continued operation is permitted. See B.

C.

D.


Lesson Plan Objective: IRMSYSE006 Given a scenario of applicable operating conditions: a.

b.

c.

Choose those sections which are applicable to the IRM system Evaluate IRM operability and determine required actions based upon system operability Determine the bases for those tech spec items associated with the IFW System IAW HCGS Technical Specifications. (SRO Only)



Reference(s): NOH05000014C - Intermediate Range Monitoring System

KA: 2 15003.G2.1.12 Ability to apply Technical Specifications for a system. (CFR: 43.2 / 43.5 145.3)

HoDe Creek 2007 ILT Exam



0

The plant is operating at 75% power. Confirmed seal failures have occurred on the “B” Recirculation Pump. The pump has just been tripped.

Which of the following describes the order for isolation of the “B” Recirculation Pump IAW HC.OP-AB.RE’V.0003, “Recirculation System / Power Oscillations” and what is the basis for this order?

A. Close the suction valve, then isolate seal purge, and then close the discharge valve. This ensures that that the suction valve is stroked against a minimal differential pressure.

Close the suction valve, then isolate seal purge, and then close the discharge valve. This will minimize additional overpressure damage to the seal package.

Close the discharge valve, then isolate seal purge, and then close the suction valve. This ensures that that the discharge valve is stroked against a minimal differential pressure.

Close the discharge valve, then isolate seal purge, and then close the suction valve. This will minimize additional overpressure damage to the seal package.

B.

C.

D.

A. Correct: Correct order and basis. See AB.RPV-0003 and SO-BB- 0002 Precaution 3.1.14. The suction valve is not designed to stroke against a high DP and may not fully close against high DP.

B. Incorrect: Incorrect basis. C. Incorrect: Incorrect order. D. Incorrect: Incorrect order and basis.



Lesson Plan Objective: ABRPV3E001: Recognize abnormal indicationsialarms andor procedural requirements for implementing Recirculation SystemPower Oscillations.

Reactor Recirculation System Operation, explain the bases for listed precautions and limitations.

RECIRCE013: Given procedure HC.OP-SO.BB-0002,

Source: INPO Bank (HC question 2/28/98)


Reference(s):

HC.OP-SO.BB-0002 Precaution 3.1.4 HC.OP-AB.RPV-0003

KA: 202001.A2.10 Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) A2.10 Recirculation Pump Seal Failure 3.5 3.9




All SRM’s are operable. The plant is shutdown for refueling.

The SRM shorting links are installed. Arc welding is in progress near the SRM preamp panels. An irradiated bundle is being lowered into the core in “C” quadrant. The SRM UPSCALE OR INOPERATIVE C3-CI alarm annunciates. “A” SRM indicates hard upscale. There was no change on the other SRMs.

What is the required immediate action?

A. Raise the fuel bundle up and out of the core and move it to a safe location in the spent fuel storage pool.

Terminate fuel movement in “A” core quadrant ONLY.

Terminate fuel movement in all core quadrants.

Terminate welding activities near the SRM preamp panels to stop the spurious alarms.

B.

C.

D.

______________I______________________I__-----------------------------------------


A. Incorrect: Not an immediate action. Plausible: If lowering the bundle caused the upscale, then removing it might correct the problem. Incorrect: AB.IC-0004 requires that all fuel movement be terminated. Plausible: “A” SFUU is the only upscale SRM. Other SRMs are normal.

C. Correct: Immediate action IAW AB.IC-0004. D. Incorrect: Follow-up action.

Plausible: Follow-up action.


B.


Lesson Plan Objective: IABIC04E003 From memory, recall the Immediate Operator Actions for Neutron Monitoring.

Source: Mod INPO Bank (HC Question 2123198)


Reference(s) : HC.OP-AB.IC-0004

KA: 215004.G2.4.49 2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. (CFR: 41.10 /43.2 1 45.6)




If drywell pressure and temperature lowering results in entering the UNSAFE region of the Drywell Spray Initiation Limit (DWT-P), what action is required?

A.

Due to a large reactor coolant leak, suppression chamber pressure was rapidly increasing Drywell sprays have been initiated Drywell pressure is 10 psig and lowering Drywell temperature is 3 10 "F and lowering

Secure all drywell sprays when the Drywell Spray Initiation Limit curve is reached.

B.

C. Emergency Depressurize the reactor.

D.

Secure drywell sprays at 9.5 psig drywell pressure and lowering.

Continue drywell sprays until drywell pressure approaches 0 psig.


A. B.

Incorrect: Curve is based on spray initiation, not securing spray. Incorrect: DW sprays remain in service until 0 psig. There is no indication in the stem that the RHR pump is needed for adequate core cooling. Incorrect: Curve is based on spray initiation, not whether to ED or not. DW temperatures are improving with sprays in service. Correct: Once initiated, DW sprays need only be secured BEFORE DW pressure reaches 0 psig. EOP-102, PCC-1, Retainment Override.

C .

D.



Lesson Plan Objective: E01 02PE006 Given plant conditions and access to the following curves determine the region of acceptable operation and explain the bases for the curve IAW the Primary Containment Control - Drywell Lesson Plan: a. Drywell Spray Initiation Limit b. Pressure Suppression Pressure



Reference(s): EOP-102, PCC-I, Retainment Override.

KA: 230000.A2.15 Ability to (a) predict the impacts of the following on the RHRLPCI: CONTAINMENT SPRAY SYSTEM MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) A2.15 High Containment / Drywell pressure 3.6 3.8



1 Pt(s) A Traversing Incore Probe (TIP) trace is in progess when a high drywell pressure event (> 1.68 psig) occurs due to a leak in the recirculation system. Three minutes following the event, the Reactor Operator reports the following indications on the TIP Valve Control Monitor:

“SQUIB MONITOR” lights - both extinguished “SHEAR VALVE MONITOR” lights -both extinguished “BALL VALVE OPEN” lights - both illuminated “BALL VALVE CLOSED’ lights -both extinguished

Whxh of the following describes the status of the TIP system and the next required operator action(s)’?

A. The system has responded as designed. Operator action is required to fire the shear valves.

The TIP detectors may not have withdrawn. Fire the shear valves, withdraw the remaining cable and then close the ball valves.

B.

C. The TIP detectors may not have withdrawn. Withdraw the detectors and verify the ball valves close.

The system bas responded as designed. Operator action is required to close the ball valves.

D.


A. Incorrect: The system did not respond as required. The ball valve should be closed. Plausible: Could conclude that squib lights out indicate that the squib fired and the line is isolated. Incorrect: The next action is to attempt a manual withdrawal. If that fails, then firing the shear valve alone will isolate the line. Plausible: Tips may not have withdrawn is correct. The subsequent actions are plausible but not correct. Correct: The ball valve should be closed. A failure to auto retract could be the problem. The next action would be to attempt to withdraw the detectors and verify the ball valve closes. Incorrect: Did not respond as designed.

B.

C.

D.


Plausible: Could conclude that manual action is needed to close the ball valves. Lights out on shear valve could be interpreted as though the shear fired to isolate the line.


Lesson Plan Objective: TIPSOOE003 Given a labeled drawing of the TIP System controls and indications in the main control room (panel 10C607): a. b.

c.

d.

Explain the function of each indicator. Assess the plant conditions that will cause the indicator to light or extinguish. Predict the effect of each control switch on the TIP System control panels. Select the conditions for permissives required for the control switches to perform their intended function.

Source: HC Bank (456564) NRC Exam 10199


Reference(s) :

TIP Lesson Plan

KA: 215001.A2.07 Ability to (a) predict the impacts of the following on the TRAVERSING IN-CORE PROBE SYSTEM ; and @) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) A2.07 Failure to retract under accident conditions 3.4 3.7

OP-AB.ZZ-109



1 Pt(s) The Unit is operating at 100% power when an inadvertent HPCI initiation and injection occurs due to a testing error.

Wlnch one of the following describes the required immediate action?

A. Enter HC.OP-AB.RPV-0003, Recirculation System / Power Oscillations and take the appropriate actions.

Reduce reactor recirculation flow and insert rods as needed to reduce reactor power.

B.

C. Actuate Isolation Logic A and C by depressing the associated trip pushbutton.

Shut the HPCI pump discharge isolation valve HV-F007 using the Control Room hand switch.

D.


A. Incorrect: This procedure is for power oscillations related to thermal hydraulic instability. The inadvertent HPCI injection should not cause instability but rather a rise in power. Correct: This event results in a cold water injection and positive reactivity insertion. The immediate action per AB.RF’V-0001 Reactor Power for a rising power condition is to reduce reactor power as needed below the pre transient value. Incorrect: Securing HPCI is a follow-up action not an immediate action. Furthermore, the procedure directs tripping HPCI not isolating it. Incorrect: Securing HPCI is a follow-up action not an immediate action. Furthermore, the procedure directs tripping HPCI not isolating the discharge valve.


Lesson Plan Objective: ABRPVlE003 From memory, recall the Immediate Operator Actions for Reactor Power.

B.

C.

D.


Source: INPO Bank (HC Question 12/18/95)


Refaence(s): HC.OP-AELRPV-0001 Reactor Power

KA: 206000.A2.17 Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) A2.17 tHPCI inadvertent initiation: BWR-2,3,4; 3.9 4.3



1 Pt(s) The following plant conditions exist:

Reactor power is 30%

Reactor startup is in progress Mode switch is in “RUN”

“A” FRVS vent unit was declared inoperable 8 hours ago. Repair is expected to be completed in 8 hours.

The ”A” EDG capability test was completed at the end of the previous shift. As you review the paperwork you determine that the EDG must be declared inoperable due to unsatisfactory acceptance criteria.

Which ONE of the following actions is required per technical specifications?

A. The reactor must be in Hot Shutdown within 12 hours.

B. The reactor must be in Cold Shutdown within 24 hours.

C. Startup may continue provided the “A” EDG is made operable within the following 72 hours.

Suspend power ascension until the “A” EDG and “A” FRVS are declared operable. Operation at the current power level is permitted provided that the “A” EDG is made operable within the following 72 hours.

D.

____________--____-____l_____________l__-------------------------------------_-----


A. Incorrect: The Unit is not in a shutdown statement. The individual LCOs for the EDG and FRVS apply. Plausible: Could misinterpret Tech Specs and incorrectly apply 3.0.3 since more than one system is inoperable; however, this would be incorrect since the individual LCOs apply. Incorrect: The Unit is not in a shutdown statement. Plausible: 24 hour Cold Shutdown statements are a part of the FRVS and EDG LCO action statements. Correct: The “A” EDG is the limiting LCO. Continued operation is permitted for the next 72 hours. The mode switch is already in run, so the startup may continue since raising power is not an OPCON change as described in 3.0.4.

B.

C.


D. Incorrect: Tech Specs do not restrict power ascension in this instance. Plausible: LCO 3.0.4 restricts OPCON changes when equipment is required to be operable in the next OPCON and there is a shutdown action associated with the LCO. In this case, the mode switch is already in RUN and raising power alone is NOT considered to be an OPCON change.


Lesson Plan Objective: TECSPCEOlO Given specific plant operating conditions and a copy of the Hope Creek Generating Station Technical Specifications, evaluate planthystem operability and determine required actions (if any) to be taken. (SRO/STA Only)

Source: INPO Bank (HC Question 0811011998) Replaced distracter (D) with a new distracter. Modified distracter (A) to change time from 24 to 12 hours to Hot Shutdown since this would make LCO 3.0.3 plausible.


Reference( s): Tech Specs (3.03, 3.0.4, EDG, and FRVS)

KA: 264000.G2.2.22 2.2.22 Knowledge of Limiting Conditions for Operations and safety limits. (CFR: 43.2 / 35.2) 4.1




The Unit was shutdown on the previous shift to begin a refueling outage. Reactor temperature is currently 185°F with the reactor head installed and fully tensioned. The “A” EDG is out of service for modification work. The MODE switch has been placed in STARTUPkIOT STANDBY for RPS testing.

What of the following describes the current Operational Condition and related requirements?

A. Startup/Hot Standby and operation in this condition is allowed provided control rods are verified to be fully inserted.

Cold Shutdown and operation in this condition is allowed provided control rods are verified to be fully inserted.

Cold Shutdown and all requirements for being in Operational Condition 2 must be met.

B.

C.

D. Startupmot Standby and all requirements for being in Operational Condition 2 must be met.


A. Incorrect: StartupIHot Standby is a mode switch position and not an Operational Condition. OPCON 2 is the “Startup” OPCON and transitioning from OPCON 4 to OPCON 2 is NOT permitted with an EDG out of service. Plausible: The mode switch is permitted to be placed in StartuplHot Standby while in OPCON 4 (Cold Shutdown) if all rods are verified to be hlly inserted. Correct: The Unit is in OPCON 4 (Cold Shutdown). The mode switch may be placed in Startupmot Standby for testing in this OPCON provided that the control rods are verified to remain fully inserted. Incorrect: The requirements of OPCON 2 are NOT required to be met before placing the mode switch in Startupmot Standby to support RF’S testing.

B.

C.


Plausible: The OPCON is correct (OPCON 4 Cold Shutdown). Incorrect: StartupIHot Standby is a mode switch position and not an Operational Condition. The requirements of OPCON 2 are NOT required to be met before placing the mode switch in Startupmot Standby to support RPS testing. Plausible: StaxtupHot Standby is the correct mode switch position for OPCON 2.

D.

Level: SRO Exam

Lesson Plan Objective: TECSPCEOOS Define or discuss the terms contained in Section 1 .O of Hope Creek Generating Station Technical Specifications.

Source: HC Bank (Q53260) 1997 NRC Exam


Reference(s): Tech Specs

KA: G2.1.22 2.1.22 Ability to determine Mode of Operation. (CFR: 43.5 / 45.13) IMPORTANCE RO 2.8 SRO 3.3



1 Pt(s) The reactor is operating at 100% power. The licensed operator "at the controls" is permitted to go to:

A. the shift clerk area to make a copy of the FXVS Monthly Surveillance Test.

B. the Neutron Monitoring System back panels (lOC608) to second verify red blocking tags.

the SM office to brief the SM on plant conditions during a plant fire.

C.

D. the RM-I1 to respond to a radiation monitor alarm.


A. Incorrect: SeeD. B. Incorrect: SeeD. C. Incorrect: SeeD. D. Correct: IAW NC.NA-AP.ZZ-0005 Att #1 the RM-I 1 is the only

area considered "At the Controls". Also reference HC.OP-AP.ZZ- 0005.


Lesson Plan Objective: ADMPR05CE004 Given a copy of the control room layout Identify the area denoted "at the controls," IAW NC.NA-AF'.ZZ-0005

Source: HC Bank (Question 561 86)


Reference(s): NC.NA-AP.ZZ-0005 Att #I HC.OP-AP.ZZ-0005

KA: G2.1.6 2.1.4 Knowledge of shift staffing requirements. (CFR: 41.10 / 43.2) IMPORTANCE RO 2.3 SRO 3.4



1 Pt(s) In accordance with site administrative processes, a lOCFR 50.59 Evaluation determines if a proposed change, test or experiment requires:

A.

B.

C.

NRC notification prior to performance.

Inspection by the NRC during the activity.

Prior NRC approval via a license amendment.

D. Evaluation for compliance with NRC regulations.

__ ~~~~~~~~ ~~~~~~~~


A.

B. C.

D.

Incorrect: 50.59 evaluates for license amendment requirements not notification requirements. Incorrect: NRC determines inspection requirements. Correct: The evaluation determines if a license amendment is required. Incorrect: Does not evaluate compliance with NRC regulations.

Level: SUO Exam CFR 55.43@)(3)

Lesson Plan Objective: ADMPROE078 From Memory State the purpose of 10CFR50.59 Reviews and Safety Evaluations. IAW NC.NA-AP.ZZ-0059.



Reference(s): LS-AA-104

KA: (32.2.5 2.2.5 Knowledge of the process for making changes in the facility as described in the safety analysis report. (CFR: 43.3 / 45.13) IMPORTANCE RO 1.6 SRO 2.7




An I&C Technician is in the middle of SRM "A" Channel Functional Test The next section of his procedure contains several discrepancies

Whxh one of the following changes is PROHIBITED as an "On The Spot Change" to the procedure?

A.

B.

C.

D.

Raising the trip setpoint tolerance to reduce nuisance alarms.

Adding clarifying remarks to a procedural step.

Changing a step which corrects a switch nomenclature error.

Adding a supervisory review signoff.


A. Correct: Raising the tolerance of the trip setpoint is a change of intent because it is not being performed to align with Technical Specifications. Incorrect: Clarifying a step is permitted under Attachment 1. Incorrect: Correcting typographical errors is permitted under Attachment 1. Incorrect: Changing the level of oversight is permitted IF the change results in increased oversight under Attachment 1.


B. C.

D.

Lesson Plan Objective: ADMPROE002 From Memory Describe what requirements must be satisfied to make an On-the-Spot change, and the required approval signatures. IAW NC.NA-AP.ZZ-0001 and NC.DM-AP.ZZ-0002.

Source: HC Bank (468868) NRC Exam 2002

Level of knowledge: M e m o q


Reference(s): NC.NA-AP.ZZ-0001 and NC.DM-AP.ZZ-0002

KA: G2.2.11 Knowledge of the procedure for controlling temporary changes. (CFR41.10i43.3i45.13) IMPORTANCERO2.5 SR03.4



1 Pt(s) Radiation Protection technicians have surveyed the Rehe1 Floor Reactor Head Laydown Area during an outage and obtained the following results:

Airborne Concentration: 0.15 DAC Area Dose Rates one foot from the source: 72 m r h

Smear Results: 750 dpmil00 cm2 gamma

Based on these results the area should be posted as a:

I. Radiation Area 11. High Radiation Area Ill. Very High Radiation Area IV. Contaminated Area V. Airborne Radioactivity Area

A. I,IV,andV

B. 1,andV

C. 111, andIV

D. I1,andN


A.

B.

C.

Incorrect: Not a Contamination Area since a Contamination Area= >lo00 dpmi100crn2. Correct: Radiation area = >5 mRem/hr to 100 mRem/hr. Airborne Radioactivity area = >IO% or 10 DAC. Incorrect: Not a Very High Radiation Area since a Very High Radiation Area = 2500 Radsmr. Not a Contamination Area since Contamination Area= >lo00 dpm/100cm2. Incorrect: Not a High Radiation Area since < 100 mRem/hr. Not a Contamination Area since Contamination Area= >I 000 dpdlOOcm2.


D.


Lesson Plan Objective: NOH04ADM024C-02 From Memory State the definition of the following terms: a. Contaminated Area b. High Radiation Area c. Locked High Radiation Area d. Radiation Area e. Restricted Area E Very High Radiation Area g. Airborne Radioactivity Area h. Declared Pregnant Woman (DPW) i. Total Effective Dose Equivalent (TEDE) IAW NC.NA-AP.ZZ-

0024

Source: HC Bank (476884) 2002 LSRO Exam


Reference( s) : NC.NA-AP.ZZ-0024

KA: G2.3.1 Knowledge of 10 CFR: 20 and related facility radiation control requirements. (CFR: 41.12 / 43.4. 45.9 / 45.10) IMPORTANCE RO 2.6 SRO 3.0



1 Pt(S) Given the following:

A Site Area Emergency was just declared 20 minutes ago due to a primary system line break that is discharging to the environment and it CANNNOT be isolated from the control room. TSC & EOF personnel are beginning to arrive at their facilities; however, these facilities have NOT been fully activated at this time. Operators are developing a plan in the OSC to manually isolate the line in order terminate the release. An Emergency Dose Authorization will be needed to isolate the line. The Radiological Assessment Coordinator (RAC) has arrived in the TSC. The Emergency Duty Officer (EDO) has NOT arrived in the TSC, and the ED0 CANNOT be reached by phone.

s

Who can authorize the Emergency Exposure in the absence of the ED0 and what is the Planned Emergency Exposure Limit (PEEL)?

A.

B.

C.

The Shift Manager is empowered to authorize up to 25 REM.

The Shift Manager is empowered to authorize up to 75 REM.

The RAC is empowered to authorize up to 25 REM.

D. The RAC is empowered to authorize up to 75 REM.


A. Correct: NC.EP-EP.ZZ-0304 states that the Shft Manager has the responsibility to authorize Emergency Exposures until the ED0 assumes his or her responsibilities. The Planned Emergency Exposure Limit is 25 REM for accident mitigation and isolating the line is an accident mitigation action. Incorrect: 75 REM is the PEEL limit for saving a life. Isolating the line is NOT a life saving action. Incorrect: The RAC is NOT authorized to grant permission for an Emergency Exposure. Incorrect: The RAC is NOT authorized to grant permission for an Emergency Exposure. 75 REM is the PEEL limit for saving a life. Isolating the line is NOT a life saving action.

B.

C.

D.


Level: SRO Exam CFR 55.43@)(4) & (5)

Lesson Plan Objective: ?’???

Source: New


Reference(s): NC.EP-EP.ZZ-0304, ‘‘Operational Support Center (OSC) Radiation Protection Response”

KA: G2.4.38 2.4.38 Ability to take actions called for in the facility emergency plan / including (if required) supporting or acting as emergency coordinator. (CFR. 43.5 / 45.11) IMPORTANCE RO 2.2 SRO 4.0




You have received an authentic notification that a large commercial airliner departed from Philadelpha International Airport a few minutes ago and then diverted from its assigned flight plan. The airliner is NOT responding to Air Traffic Control. The airliner is flying at low altitude at about 500 feet above ground level and is heading towards the Sa ldHope Creek complex and it is closing at 250 mph. The airliner has been determined to be a credible threat to the station and it is less than 12 minutes away.

Which one of the following plant page scripts is correct for this situation?

A. Attention all personnel, Attention all personnel. A credible security threat exists at the station. Communicators report to the Control Room. AU onshift personnel report to the OSC. AU onsite ERO responders report to the NOSF immediately. All other personnel place your work area in a safe condition; exit the protected area immediately and proceed home.

B. Attention all personnel, Attention a11 personnel. A credible security threat exists at the station. Communicators report to the Control Room. All other personnel place your work area in a safe condition; take immediate cover and shelter in place until further notice.

C. Attention all personnel, Attention all personnel. A credible security threat exists at the station. All ERO members immediately report to your emergency response facility. All other personnel place your work area in a safe condition, exit the protected area immediately and proceed home.

D. Attention all personnel, Attention all personnel. A credible security threat exists at the station. Primary communicator report to the Control Room. All on shift personnel and onsite ERO responders report to the NOSF immediately. All other personnel place your work area in a safe condition; take immediate cover inside the nearest buildiog and away from the containment building.



A.

B.

C.

D.

Incorrect: This is the script for an airborne threat when there is greater than 15 minutes lead time. Incorrect: This is the script for a NON airborne security threat with less than 15 minutes lead time. Incorrect: This is the script for a NON airborne security threat with greater than 30 minutes lead time. Correct: Th~s is the correct script for an imminent airborne threat with less than 15 minutes lead time. See NC.EP-EP.ZZ-0102.



Source: New


Reference(s) : NC.EP-EP.ZZ-102

KA: G2.4.43 2.4.43 Knowledge of emergency communications systems and techniques. (CFR: 45.13) IMPORTANCE RO 2.8 SRO 3.5


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